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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Estimating Radionuclide Activity in Surface Contamination for GTCC Activated Metal Waste Streams - 18236

The U.S. Nuclear Regulatory Commission's licensing requirements for land disposal of low-level radioactive waste (LLRW) in 10 CFR Part 61 classify LLRW wastes as Class A, B, C, or Greater-Than- Class C (GTCC). GTCC waste includes activated metals that are generated primarily by nuclear power reactors. Activated metals represent a significant volume (880 cubic meters) of GTCC wastes that may be generated as current commercial power reactors are decommissioned. Additionally, these wastes are expected to contain radionuclides that are very long-lived and are mobile in the environment (e.g., Tc-99 and I-129) as well as transuranic radionuclides that, while not as mobile as other radionuclides, have greater potential for health consequences should such radionuclides reach receptor locations in sufficient concentrations. Quantities of surface contamination on activated metals have typically been reported using a combination of assumptions, calculations, and measurements. For example, projections of future activity associated with surface contamination, or the surface activity, for the activated metals is dependent on assumptions made regarding fuel behavior (e.g., modern fuel performance is expected to be different from fuel performance in the 1970's). In this study, we examined published information to better understand how assumptions, calculations, and measurements used for estimating quantities of surface contamination on activated metals can influence the estimated quantities. The analysis in this report was performed to better understand how estimates for the activated metal inventory of GTCC metal could influence potential decisions regarding the siting and design of disposal facilities. It was found that the variability in surface contamination either measured or estimated was sufficiently large that characterization of the radionuclide inventory for GTCC activated metals, especially the transuranic radionuclides, is critically important for establishing the siting and design requirements for a disposal facility. Overly conservative assumptions for the inventory could significantly limit disposal options, whereas, overly optimistic assumptions with respect to the inventory could lead to a disposal facility that may not protect off-site receptors. (authors)

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DOE OSTI2018

Containing Tc-99 Using a Multi-sorbing Barrier Material - 18249

Immobilization of Tc-99 is a challenge at on-site waste disposal facilities in the DOE complex because the oxyanion pertechnetate (TcO{sub 4}{sup -}) binds weakly (if at all) to the surface of mineral solids and other conventional barrier materials (distribution coefficient, Kd<2 mL/g). The study described herein evaluated TcO{sub 4}{sup -} uptake from aqueous solutions using a multi-sorbing organo-clay barrier material synthesized from sodium-bentonite (Na-B) exchanged with hexadecyltrimethylammonium (HDTMA). The exchange transformed the negatively charged surface of Na-B to a positively charged surface to bind TcO{sub 4}{sup -}. Batch tests were conducted to evaluate sorption of TcO{sub 4}{sup -} onto the multi-sorbing clay barrier material using ammonium-pertechnetate solutions with activity ranging from 1000 to 2,000,000 pCi/L. Tests were conducted with a solid-to-liquid ratio of 1:100 in 50-mL centrifuge tubes rotated at 30 rpm for 24 hr. Solutions were created by dissolving ammonium-pertechnetate in DI water or synthetic low-level radioactive waste leachate. Solids were separated from the test solution by centrifugation at 4000 rpm for 10 min. A liquid scintillation counter was used to measure the Tc-99 in solution. Isotherms from the batch tests show that the multi-sorbing barrier material is very effective in removing Tc-99, with Kd > 2900 mL/g. Using this high capacity sorbent in engineered barriers may provide a sustainable solution to control release of Tc-99 and other anions of concern from on-site disposal facilities. (authors)

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DOE OSTI2018

Durable Tellurite Glasses for Vitrification of Technetium Wastes - 18270

Technetium-99 (Tc-99) can bring a significant environmental risk because it has a long half-life and high fission yield. The present work has developed alumino tellurite glass to immobilize Tc wastes generated by the pyrochemical processing technology. Rhenium (Re) was used as a non-radioactive surrogate for Tc. The Re{sub 2}O{sub 7} concentration in the tellurite glasses was as high as 9.11 mass% with retention of 86% on average. X-ray absorption spectroscopy analysis revealed that Re exists as Re{sup 7+} and is surrounded by 4 oxygen atoms. The 7-day PCT results were satisfied with the immobilized low-activity waste regulatory limit with Ca(ReO{sub 4}){sub 2} content less than 12 mass%. The long-term dissolution test was conducted for 256 days and a rate drop has occurred in which the rate of leaching gradually decreased. An alteration layer depleted soluble element was observed on the leached glass surface. These results suggest that tellurite glass can be a good candidate as a matrix for immobilizing Tc. (authors)

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DOE OSTI2018

Technetium-99 Behavior in Low-Temperature, Low-Pressure Evaporation Systems at the Hanford Site - 18461

This paper seeks to provide a technical basis for predicting the behavior of Technetium-99 (Tc-99) in the Effluent Management Facility (EMF) evaporator during the Direct Feed Low-Activity Waste (DFLAW) mission at the Hanford Site. This is done by using the 242-A Evaporator and the DM10 Melter System evaporator as analogues for the predicted design and operating strategy of the EMF evaporator. The focus of this investigation is to establish how Tc-99 behavior coincides with the behavior of several process variables (e.g. feed Tc-99 concentrations, feed sources, extent of evaporative concentration) associated with low-temperature, low-pressure evaporator operations. The primary measure of Tc-99 behavior was the partitioning of Tc-99 from the feed stream into the condensate stream (i.e. boil-off) of the evaporators. Analysis of 242-A and DM10 Melter System evaporator data allowed for the identification of the following trends regarding Tc-99 behavior in low-temperature, low-pressure evaporator systems: Tc-99 partitions minimally from the feed to the condensate stream and Tc-99 partitions consistently across a variety of operating and feed conditions. Given the similarities in design and operating strategies between the studied evaporator systems and the EMF evaporator, it is reasonable to suggest that these trends can be used to aid in planning for future EMF operations. (authors)

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

Dynamic Leaching Method for Intact Salt-stone Samples - 18495

At the Savannah River Site (SRS) low activity salt solution is stabilized via encapsulation within a grout termed salt-stone. Salt-stone is emplaced into large (multi-million liter) concrete storage containers referred to as Salt-stone Disposal Units (SDUs). Technetium-99 (Tc-99) and iodine-129 (I-129) are long-lived radionuclides contained in the low activity salt waste and subsequently incorporated into the grout waste form: both radionuclides are contributors to the potential long-term radiation exposure of the public and environment surrounding the SDUs. With respect to Tc-99, in the reducing, high pH environment within the grout, Tc-99 is expected to be relatively immobile since it exists in a reduced Tc(IV) oxidation state in the form of sparingly soluble hydrated oxides (TcO{sub 2}.xH{sub 2}O) or sulfides (TcS{sub x}). However, in the presence of O{sub 2} (associated with the future infiltration of air or oxygenated ground waters into the salt-stone monolith) it is possible for redox-sensitive Tc(IV) to transition into highly soluble (and mobile) Tc(VII) species, such as pertechnetate (TcO{sub 4}{sup -}), which are more readily transported to the surrounding environment. In contrast, I-129 is highly soluble and its leaching behavior is not anticipated to be significantly impacted by potential changes in redox environment or pH within salt-stone. Traditional approaches to quantifying the leaching behavior of Tc-99 and I-29 from cementitious matrices have involved partitioning experiments using size-reduced (crushed/ground) samples, and determination of the radionuclide fraction immobilized by the cementitious solids. Such experiments create artificially high solid-liquid contact areas that likely result in higher Tc-99 and I-129 leachate concentrations than would be expected for intact, monolithic samples. In the current study a new technique, termed the Dynamic Leaching Method (DLM), is being used to initially investigate the Tc-99 leaching behavior from monolithic salt-stone samples; the system will be used at a later date to evaluate the leaching characteristics of I-129. The data derived using this technique is intended to inform the SRS Salt-stone Disposal Facility (SDF) Performance Assessment (PA) which models the long-term transport of radionuclides from the SDUs to the environment. The DLM utilizes a flexible-wall permeameter to achieve saturated leaching under an elevated hydraulic gradient to simulate and accelerate the future transport of groundwater through salt-stone. Leachates are regularly analyzed with respect to radionuclide concentrations, key constituent concentrations (e.g., Al, Ca, Mg, Na, Si, CO{sub 3}{sup 2-}, NO{sub 2}{sup -}, NO{sub 3}{sup -}, SO{sub 4}{sup 2-}), pH, and reduction potential (E{sub h}) in order to develop an understanding of how chemical changes associated with the pore solution influence grout weathering and radionuclide leaching behavior. This paper provides a brief update with respect to data derived from preliminary Tc-99 leaching evaluations as the DLM was being developed, but the paper is predominantly focused on new sample data in which the DLM technique, pre- and post-DLM salt-stone sample characterization, and leachate analysis protocols have been optimized. (authors)

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DOE OSTI2018

Five Years of 200 West Pump and Treat Facility Operation Success - 18519

The 200 West Pump and Treat (P and T), on the Department of Energy's (DOE) Hanford Site, in Richland, Washington has a 9,464 liters per minute capacity designed to capture and treat an 8 km{sup 2} contaminated groundwater plume and reduce the mass of contaminants of concern (COCs) by 95 percent in 25 years. Treatable contaminants include both radioactive and non-radioactive compounds. After five years of continuous operation, the 200 West P and T treated over 17 billion liters of contaminated groundwater and removed over 337,000 kg of nitrate, 12,800 kg of carbon tetrachloride, 180 kg of uranium, and 10 curies of technetium-99 activity. Thirty one extraction wells strategically installed across the plumes pull contamination in from surrounding areas directing contaminated groundwater to the 200 West P and T for treatment. The 200 West P and T successfully removes contaminants to below cleanup levels which are at (or lower than) the drinking water standard. Twenty eight injection wells return treated water to the aquifer providing flow-path control to mitigate migration of contamination toward the Columbia River, a source of drinking water to the communities downstream. Contaminant concentrations in most monitoring wells are declining and the remedy is on target for accomplishing the mass removal goal ahead of schedule. In 2015, the 200 West P and T was expanded to include treatment for uranium from other operable units. With the uranium treatment expansion, the 200 West P and T is capable of cost-effectively treating contaminated groundwater beneath the 194 square kilometer (km{sup 2}) Central Plateau. (authors)

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DOE OSTI2019

Pump and Treat Impacts on Aquifer Conditions and the Development of System-wide Performance Strategies - 19077

The 200 West 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 (OU). Initially designed to support the 200-ZP1 Remedial Action Objectives, the use of the P and T system has evolved over time to address contaminants from other regions of the aquifer as well. The P and T system currently receives water from other regions within the Central Plateau that are biogeochemically distinct, treating groundwater with elevated concentrations of carbon tetrachloride, cyanide, uranium, technetium-99, and nitrate. Given the varying geochemical signatures from current and potential future sources, combined with the operational conditions and amendments used for treatment processes, there is a need to identify the impacts of the P and T system on the aquifer and aquifer impacts on P and T performance. To this end, initial laboratory analyses and geochemical modeling efforts focused on characterizing the baseline geochemical environment for the aquifer in the 200-ZP-1 OU. Several sediment characterization techniques, including geologic descriptions, water and acid extractions, and microbiological measurements were selected to determine key geochemical and microbiological constituents for the baseline, prior to injection of the effluent from the treatment plant. Different geochemical components within sediment samples from the aquifer were identified that may significantly affect how the sediments react to P and T injection water. Batch experiments were also used to identify aquifer impacts on operational performance and support development of some column studies. These experiments were designed to test and evaluate variable P and T effluents using actual effluent solutions from the treatment plant as well as simulated Hanford groundwater at different temperatures to identify the important geochemical and biological reactions that might occur within the aquifer. Initial results were used for geochemical modeling of the effluents and batch experiments. These results, along with the baseline data obtained from the initial sediment characterization, are intended to feed into the design of the column studies and inform conceptual models and modeling efforts for evaluating the fate and transport of introduced contaminants and co-injectates, and their short- and long-term impact on the aquifer-system injection capacity. This evaluation is designed to provide additional information on the major geochemical and microbiological reactions that are driven by the injection of the P and T effluent into the aquifer. Reactive transport modeling of the laboratory experiments and extrapolation to a larger-scale modeling domain provide information that can be used for 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 OUs on the Hanford Site Central Plateau. (authors)

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DOE OSTI2019

Analysis of Tc and U Spent Ion Exchange Resins from Hanford 200 West P and T Facility - 19202

The 200 West Area Pump-and-Treat (P and T) facility at the Hanford Site is being used to remove radionuclides such as uranium (U) and technetium-99 (Tc-99) from the contaminated groundwater. To support evaluation of U and Tc ion exchange (IX) treatment operational conditions and performance with various influent conditions, knowledge of resin capacity and contaminant retention is needed. To meet these objectives, samples of the spent Dowex{sup R} 21K and Purolite{sup R} A530E IX resins from the 200 West P and T facility used for the treatment of U and Tc, respectively, have been obtained and analyzed for the target analytes and other retained contaminants. The objectives of this work included 1) determination of the effective U and Tc loading on the resins, 2) evaluation of redox and chemical speciation of U and Tc retained by the resins and impact of the feed composition on the retention of U by the resin, and 3) quantification of other contaminants and groundwater constituents retained on the resins. Speciation of U and Tc-99 on the resin was determined by a combination of X-ray absorption fine structure, nuclear magnetic resonance, X-ray photoelectron spectroscopy (XAS), and optical spectroscopic techniques. Nondestructive analytical techniques including gamma energy analysis and low-energy photon spectroscopy were incorporated to quantify radionuclides retained on the resin. Scanning electron microscopy and energy-dispersive X-ray spectroscopy were used to observe the IX beads and quantify chemical constituents on the surface and penetration through the beads. Destructive analysis of the resin was explored through wet oxidation to identify and quantify co-contaminants sorbed to the resins. Inductively coupled plasma-optical emission spectroscopy (ICP-OES) and inductively coupled plasma-mass spectrometry (ICP-MS) were used to analyze for metals, iodine, and specific isotopes retained on the resin, including Tc-99 and U-238. A single pertechnetate Tc-99 species was identified to be retained on the resin. For U, the obtained structural information was correlated with the laboratory findings using pristine resin as reference material to identify the nature of the U complexes and predict U retention by the resin as a function of variable influent conditions. Both resins demonstrated high loading with sulfate, silica, calcium, iron, zinc, and other groundwater constituents, suggesting that the competing anions, fouling, or a combination thereof significantly reduced their effective capacity for U and Tc-99. A significant amount of nonradioactive iodine-127 (I-127) and long-lived radioactive iodine-129 (I-129) isotope was also retained on the resins. Obtained quantitative information provided insights on the impact of water chemistry and the presence of common mineral constituents on the retention of contaminants of concern, including U, Tc-99, and I-129. (authors)

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DOE OSTI2019

Using Correlations to Cs-137 to Reduce Uncertainty in SRS Tank Farm Tc-99 and I-129 Inventories - 19266

Performance Assessment (PA) modeling of the Saltstone Disposal Facility (SDF) shows that iodine-129 (I-129) dominates peaks of potential future doses, followed by contributions from technetium-99 (Tc-99). While these radionuclides are important from a PA perspective, they are relatively unimportant from a safety analysis or occupational exposure perspective. As such, the storage of Tc-99 and I-129 within liquid waste at Savannah River Site (SRS) have not been well tracked, resulting in considerable uncertainty in the estimated inventories of Tc-99 and I-129 used for PA modeling. Alternatively, cesium-137 (Cs-137) is considered a high-risk radionuclide from an occupational exposure perspective and is well characterized. Within the liquid waste at SRS, analyses have shown that concentrations of Cs-137 are closely correlated to both Tc-99 and I-129. These correlations and the abundance of Cs-137 data were leveraged to significantly reduce inventory uncertainty for waste destined to be disposed at the SDF. Previous attempts to mathematically define these correlations had two flaws: (1) the defined correlations failed to address the impact of the relatively short half-life of Cs-137 compared to the long half-lives of Tc-99 and I-129, and (2) the analyses relied on a small number of samples. To address the issue of the decay rate from the short half-life, all Cs-137 concentrations were decayed to a single future date. To improve the number of samples used to develop the correlations, a literature review was performed resulting in a larger data set. The data underwent a validation process, then the correlations were updated. The resulting correlations were used to improve inventory estimates for Tc-99 and I-129 for an ongoing revision to the SDF PA. (author)

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DOE OSTI2019

Performance Assessment for Closure of the Calcined Solids Storage Facility at the INL Site - 19289

The Calcined Solids Storage Facility is composed of six discrete stainless steel 'bin sets' at the Idaho Nuclear Technology and Engineering Center at the INL Site. Approximately 4,440 m{sup 3} of the dry, granular calcine waste is stored in the facility. Closure of the facility will entail removal of the calcine waste followed by grouting. A 5-cm thick residual amount of calcine was assumed to remain at the bottom of the bins. Important radionuclides in the residual layer for the groundwater pathway were identified as the fission and activation products Tc-99, Se-79, I-129, Cs-135, H-3, and actinides Pu-239, and Pu-240. The stainless-steel bins prevent near-term release of grouted residual calcine waste to the environment. Future releases were conservatively based on the corrosion rate of the weld seams on the stainless-steel bins, which site-specific studies measured mean tunnel corrosion rates of 0.0132 mm/yr with a coefficient of variation of 0.333. Based on tunnel corrosion rates, mean bin hydrologic containment times were estimated to be between 238 years and 1,191 years, depending on the thickness of the bin. Any breach in the weld seam was assumed to result in water flow and radionuclide leaching through the bin. The structural integrity of the bin was assumed to be controlled by the general stainless-steel corrosion rate, which was much slower (3.5 x 10{sup -5} mm/yr), resulting in structural integrity times between 24,000 years to over 100,000 years. A mixing-cell model was used to compute water flow and transport through the hydrologically failed bins and the vadose zone. Credit was taken for the geochemical properties of the grout used to fill the bins at closure, which significantly slows the releases from the bins for key radionuclides like Tc-99. A semi-analytical model for three-dimensional dissolved-phase transport in an aquifer of constant unidirectional flow, infinite length and width, and finite thickness was used to compute concentrations in the aquifer across the model domain for a time varying source. Cumulative impacts from all bin sets were calculated by summation across the bin set sources that were superimposed on a uniform Cartesian grid. Effective doses were calculated for direct groundwater ingestion from wells located 100-m downgradient from each of the bin sets and ingestion of meat, milk, and vegetables irrigated with water derived from the well. Intruder doses were based on a well drilled through the grouted waste and cuttings spread around a home site after a 500-year institutional control period. Peak doses during the 1,000-year compliance period were less than 5% of the 250 μSv/yr (25 mrem/yr) performance objective and intruder doses were generally less 5% of the of the 5000 μSv/yr (500 mrem) and 1000 μSv/yr (100 mrem) performance objective for acute and chronic intruder exposure, respectively. Groundwater concentrations were all less than the maximum contaminant limits. In all cases, doses were highest for the oldest bin set because, compared to the other bin sets, it had the highest radionuclide concentrations in the calcine waste. The performance assessment results indicate that performance objectives can be met leaving considerably more than 5-cm of residual calcine waste in the bin sets. (authors)

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DOE OSTI2019

Effect of Zero Valent Iron Pre-treatments on Reductive Removal of Aqueous Tc-99 - 19379

Zero-valent iron (ZVI) is being investigated as a potential material for technetium-99 (Tc-99) reductive separation from the secondary liquid waste, i.e. off-gas condensate that will be generated during vitrification of the law activity waste (LAW) at the Hanford site. In this work, we performed a comparative analysis of Tc(VII) removal kinetics via its reduction with the commercially available ZVI materials varying in particle size (granules, filings, and 200 mesh). In addition to particle size differences, reduction removal potential was also evaluated as a function of as-received ZVI material versus a surface treatment. ZVI treatment experiments with 0.4 M HCl and 1 M NaCl were conducted to study the effect of iron oxide/oxyhydroxide formation on its reductive efficiency for aqueous Tc(VII) removal. Scanning electron microscopy (SEM) with energy dispersive spectroscopy (EDS) analyses were conducted for all ZVI materials in the presence and absence of treatments to evaluate morphological differences (SEM) and chemical composition (EDS) of the materials. Reductive removal of Tc(VII) was also analyzed using a first-order kinetic model. The results of the study indicated that rate constants did not change with increased oxide/oxyhydroxide content in large ZVI materials (granules and filings), but decreased significantly for the relatively small ZVI (200 mesh), treated with 0.4 M HCl and 1 M NaCl. This study demonstrates that the commercial ZVI material 200 mesh, used as received, is a promising candidate for the removal of Tc from liquid secondary waste generated during Hanford LAW vitrification. (authors)

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DOE OSTI2019

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

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

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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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

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

Development of a Conditioned System-Level Groundwater Model to Evaluate Long-Term Groundwater Impacts within a Performance Assessment - 20115

A preliminary performance assessment (PA) of single-shell tank Waste Management Area (WMA) A-AX located at the U.S. Department of Energy (DOE)'s Hanford Site in southeastern Washington is being conducted to satisfy the requirements of DOE Order 435.1 [1] as it relates to closure of the single-shell radioactive waste tanks in the WMA A-AX tank farms. A PA assesses the fate, transport, and impacts of radionuclides within a low-level radioactive waste disposal facility in its assumed closure configuration and the subsequent potential doses to humans over a 1,000-year compliance period and a 10,000-year performance evaluation period. The WMA A-AX preliminary PA evaluation is structured around the complementary use of process-level and system-level models to calculate the facility performance against established DOE Order 435.1 [1] performance objectives. Process-level models are those that represent a detailed phenomenological representation of processes of concern in the PA. Process models typically only represent one or a few of the components of the PA, such as groundwater flow and transport, and must be integrated with other modeling elements to perform PA calculations. System-level models are those that are abstracted from the process models, retaining the essential features of the process model, while allowing integration of all aspects of the PA in a single modeling framework. System-level models are often characterized by coarser numerical discretization, lower dimensionality, or other similar simplifications compared to the process-level model. Traditionally, a three-dimensional (3-D) process-level model is utilized primarily to evaluate the long-term impact on groundwater and the potential doses to individuals who consume contaminated groundwater. System-level models are also utilized to evaluate the groundwater pathway in PAs. These models typically have reduced dimensionality (1-D) and are conditioned utilizing flow fields (Darcy fluxes) and moisture content distributions that are abstracted from the process level models. The abstraction approach assures that the flow field in both models is consistent for a specific set of input parameters for flow, differing only in the discretization and dimensionality of the two models. The preliminary WMA A-AX PA incorporates a detailed representation of the geological system and hydraulic properties within the 3-D model STOMP{sup C} numerical code so that the effects of relevant features and processes on water flow and radionuclide transport in the subsurface can be evaluated. The complementary system-level model is developed utilizing the GoldSim{sup C} code to implement a simplified, 1-D equivalent model to represent the groundwater pathway. Contaminant transport through the vadose zone and unconfined aquifer for Tc-99 and I-129 were evaluated in each of the models. Adjustments to the saturated portion of the GoldSim{sup C} 1-D model were required to mimic the dispersion effect captured with the 3-D STOMP{sup C} model. Once this conditioning was conducted, highly similar results for the transport of Tc-99 and I-129 were achieved at the point of calculation, located 100 m downgradient from the WMA A-AX fenceline. These radionuclides represent elements that are regarded as primary dose drivers in the PA groundwater pathway analysis. The high degree of conformance between the two models suggests that the use of the equivalent 1-D system model is suitable for evaluating the full suite of radionuclides that will be released from the tank sources within WMA A-AX over the 1,000-year compliance period and over the 10,000-year evaluation period. (authors)

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DOE OSTI2020

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

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DOE OSTI2020

Investigation into Reductive Separation of Tc-99 from Low Activity Waste Off-Gas Condensate by Zero Valent Iron Materials - 20311

This study is investigating the potential for reductive separation of technetium-99 (Tc) from off-gas condensate generated from vitrification of legacy tank waste at the Hanford Tank Waste Treatment and Immobilization Plant. Management of Tc is a high-priority activity for the U.S. Department of Energy (DOE) Office of Environmental Management (EM) complex, which must be addressed based on knowledge gaps and technology development needs outlined in the DOE-EM Technetium Management Program Plan [1]. Technetium is one of the main risk drivers in the permanent disposal of low activity waste (LAW) by vitrification at the Hanford Site due to its high volatility and only fractional incorporation into borosilicate glass. Volatilized Tc will be captured by an off-gas treatment system, and the current plan is to continuously recycle generated condensate back to the LAW glass melter feed. Off-gas recycling is effective at increasing Tc loading in the LAW glass, but it also disproportionately increases concentrations of sulfate, halides, and other volatile compounds impeding overall LAW processing leading to significantly greater waste glass volume requirements. Removal of Tc from secondary liquid waste offers an alternative disposal route to continuous off-gas recycling which may result in significant reduction of the volumes of LAW waste glass and substantial cost savings. The objective of this research funded by the DOE-EM Minority Serving Institutions Partnership Program (MSIPP) is to determine the suitability of zero valent iron (ZVI) as a separation technology to remove Tc from the off-gas condensate stream to minimize its recycle. While ZVI has been proposed for various remediation applications with Tc, it has not been studied for its suitability for removing Tc from the LAW off-gas condensate. The goal is to identify the optimal conditions for which Tc{sup (VII)} will be reduced to Tc{sup (IV)} species and incorporated into an iron matrix. This paper presents a programmatic overview of our evaluation of commercial ZVI materials for their efficiency towards Tc{sup (VII)} reductive removal from simplified off-gas simulant solutions under aerobic conditions. Tc concentrations in solutions were monitored for up to 30 days to determine efficiency and kinetics of reductive removal of Tc. The best performing candidate materials were selected for further experiments evaluating the effects of (1) variable ionic strength, (2) the presence of chromium and corrosive anions, and (3) ZVI aging on Tc{sup (VII)} reductive removal in batch experiments. (authors)

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