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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Ion Exchange Processing of AN-107 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System

The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. In support of the Direct Feed Low-Activity Waste (DFLAW) Program for expediting Hanford tank waste supernate treatment, laboratory-scale ion exchange processing using prototypic unit operations was conducted on AN-107 tank waste at the Pacific Northwest National Laboratory Radiochemical Processing Laboratory. This report describes the small-scale ion exchange testing with 13.7 L of diluted and filtered supernate from Tank 241-AN-107 (hereafter referred to as AN-107) at 16 °C (62 °F). One of the waste acceptance criteria (WAC) for the Waste Treatment Plant (WTP) Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the AN-107 tank waste to meet this criterion, only 0.147% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 678. Testing with AN-107 matched current Tank Side Cesium Removal (TSCR) facility prototypic operations where a lead-lag configuration was used until the lag column reached the WAC limit, then a polish column was brought online for continued processing in a lead-lag-polish column configuration. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the crystalline silicotitanate (CST) bed, matched the expected flowrate at TSCR. The Cs-decontaminated product was retained for vitrification testing (to be reported separately). The lead column reached 40% Cs breakthrough after processing ~1700 BVs of feed; the 50% Cs breakthrough was extrapolated from the breakthrough data to occur at 1873 BVs. Testing compared to previous AP-101 and AP-107 testing at 16 °C showed ~300 BV increases in volume processed to reach the WAC limit for both lead and lag columns. The increase in capacity was determined to be due to the significantly lower K concentration in the AN-107 compared to the other tank waste matrices. A comparison in breakthrough curves for the three tests indicated slightly slower kinetic behavior in the AN-107, with variations in feed matrices (high organic complexants) likely responsible for the deviation. The Cs effluent from the lag column reached the WAC limit after processing 1097 BVs. Anticipating this breakthrough point, the polish column was preemptively installed around 900 BVs. Cs breakthrough from the lag column began at 500 BVs, reaching 3.06×10 0 µCi/mL, or 2.6 % Cs breakthrough, after processing all 1700 BVs of feed. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics.

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

Special Analysis for the Disposal of the Idaho National Laboratory Idaho Chemical Processing Plant Low-Level Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This special analysis (SA) evaluates whether the Idaho National Laboratory (INL) Idaho Chemical Processing Plant (ICPP) Low-Level waste stream is suitable for disposal by shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site (NNSS). The INL ICPP Low-Level waste stream consists of debris and solidified liquids. The INL ICPP Low-Level waste stream requires an SA to resolve two unreviewed disposal questions (UDQs). The first UDQ addresses whether it is acceptable to dispose of radium-226 (226Ra) at an activity concentration that exceeds the NNSS Waste Acceptance Criteria (WAC) Action Level. The second UDQ evaluates the acceptability of disposing of five long-lived radionuclides without WAC action levels. Long-lived radionuclides without action levels are radionuclides not evaluated in the performance assessment.

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

Recovery of valuable metals from electronic waste using a novel ammonia-based hydrometallurgical process

The growing quantity of waste electrical and electronic equipment (WEEE), also known as electronic waste (E-waste), has been an area of growing public concern. The abundance of valuable metals contained in waste printed circuit boards (WPCBs) has made it a promising secondary resource, especially for Cu and Au. Although the recovery of metals from WPCBs via hydrometallurgical routes has been studied extensively over the past 20 years, most of the research has been limited in the laboratory. In current hydrometallurgical processes, strong acids and expensive oxidizers are often used to ensure a high recovery of metals without considering the sustainability aspects of the environment and economics. To improve upon current hydrometallurgical offerings, the current study seeks to develop an energy-saving, environmentally friendly, economic and sustainable process to efficiently recover the valuable metals from real-world end-of-life WPCBs. The new contributions presented in this study are 1) design and evaluation of a comprehensive hydrometallurgical flowsheet; 2) employment of real end-of-life PCBs as feed materials in an investigation on Cu-NH3 leaching kinetics; 3) further application of kinetic model on a counter flow process simulation; and 4) evaluation of the influences by co-existing metals in Au-S2O3 leaching and recommendation for favorable leaching conditions.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2011

Diffusion and Leaching Behavior of Radionuclides in Category 3 Waste Encasement Concrete and Soil Fill Material – Summary Report

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Such concrete encasement would contain and isolate the waste packages from the hydrologic environment and would act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed, and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow and/or diffusion and move into the surrounding subsurface environment. Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. The retardation factors for radionuclides contained in the waste packages can be determined from measurements of diffusion coefficients for these contaminants through concrete and fill material. Some of the mobilization scenarios include (1) potential leaching of waste form before permanent closure cover is installed; (2) after the cover installation, long-term diffusion of radionuclides from concrete waste form into surrounding fill material; (3) diffusion of radionuclides from contaminated soils into adjoining concrete encasement and clean fill material. Additionally, the rate of diffusion of radionuclides may be affected by the formation of structural cracks in concrete, the carbonation of the buried waste form, and any potential effect of metallic iron (in the form of rebars) on the mobility of radionuclides. The radionuclides iodine-129 ({sup 129}I), technetium-99 ({sup 99}Tc), and uranium-238 ({sup 238}U) are identified as long-term dose contributors in Category 3 waste (Mann et al. 2001; Wood et al. 1995). Because of their anionic nature in aqueous solutions, {sup 129}I, {sup 99}Tc, and carbonate-complexed {sup 238}U may readily leach into the subsurface environment (Serne et al. 1989, 1992a, b, 1993, and 1995). The leachability and/or diffusion of radionuclide species must be measured to assess the long-term performance of waste grouts when contacted with vadose-zone pore water or groundwater. Although significant research has been conducted on the design and performance of cementitious waste forms, the current protocol conducted to assess radionuclide stability within these waste forms has been limited to the Toxicity Characteristic Leaching Procedure, Method 1311 Federal Registry (EPA 1992) and ANSI/ANS-16.1 leach test (ANSI 1986). These tests evaluate the performance under water-saturated conditions and do not evaluate the performance of cementitious waste forms within the context of waste repositories which are located within water-deficient vadose zones. Moreover, these tests assess only the diffusion of radionuclides from concrete waste forms and neglect evaluating the mechanisms of retention, stability of the waste form, and formation of secondary phases during weathering, which may serve as long-term secondary hosts for immobilization of radionuclides. The results of recent investigations conducted under arid and semi-arid conditions (Al-Khayat et al. 2002; Garrabrants et al. 2002; Garrabrants and Kosson 2003; Garrabrants et al. 2004; Gervais et al. 2004; Sanchez et al. 2002; Sanchez et al. 2003) provide valuable information suggesting structural and chemical changes to concrete waste forms which may affect contaminant containment and waste form performance. However, continued research is necessitated by the need to understand: the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties, and the associated impact on contaminant release. Recent reviews conducted by the National Academies of Science recognized the efficacy of cementitious materials for waste isolation, but further noted the significant shortcomings in our current understanding and testing protocol for evaluating the performance of various formulations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Ion Exchange Processing of AW-105 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System

The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. In support of the Direct Feed Low-Activity Waste (DFLAW) Program for expediting Hanford tank waste supernate treatment, laboratory-scale ion exchange processing using prototypic unit operations was conducted on AW-105 tank waste at the Pacific Northwest National Laboratory Radiochemical Processing Laboratory. This report describes the small-scale ion exchange testing with 9.2 L of diluted and filtered supernate from Tank 241-AW-105 (hereafter referred to as AW-105) at 16 °C (62 °F). One of the waste acceptance criteria (WAC) for the Waste Treatment Plant (WTP) Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the AW-105 tank waste to meet this criterion, only 0.225% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 445. Testing with AW-105 matched current Tank Side Cesium Removal (TSCR) facility prototypic operations where a lead-lag configuration was used until the lag column reached the WAC limit, then a polish column was brought online for continued processing in a lead-lag-polish column configuration. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the crystalline silicotitanate (CST) bed, matched the expected flowrate at TSCR. The Cs-decontaminated product was retained for vitrification testing (to be reported separately). The lead column reached 83% Cs breakthrough after processing ~1500 BVs of feed; the 50% Cs breakthrough was interpolated from the breakthrough data and occurred at 1041 BVs. Despite the AW-105 having a significantly higher K concentration (0.55 M compared to 0.10 M), testing compared to previous AP-107 ion exchange column testing at 16 °C showed no difference in BVs processed to reach the WAC on the lead column and only an approximate ~20 BV decrease in volume processed to reach the WAC limit on the lag column. The negligible differences in capacity despite the 5x concentration differences in K was determined to be due to the significantly lower NO3 concentration in the AW-105 supernate compared to the AP-107 tank waste matrix. A comparison in breakthrough curves for the two tests also indicated slightly faster kinetic behavior in the AW-105, with the variations in feed matrices (lower NO3 concentration) likely responsible for the deviation. The Cs effluent from the lag column reached the WAC limit after processing 772 BVs. Anticipating this breakthrough point, the polish column was preemptively installed around 675 BVs. Cs breakthrough from the lag column began at 300 BVs, reaching 1.10×10 1 µCi/mL, or 14.13 % Cs breakthrough, after processing all 1500 BVs of feed. The polish column processed nominally 830 BVs and reached 2.10×10 -1 µCi/mL, or 0.27 % Cs breakthrough at the conclusion of the test. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Actual waste demonstration of the nitric-glycolic flowsheet for sludge batch 9 qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch is processable. Based on the results of this actual-waste qualification and previous simulant studies, SRNL recommends implementation of the nitric-glycolic acid flowsheet in DWPF. Other recommendations resulting from this demonstration are reported in section 5.0.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1992

Annual report

The overall goal of the Tuskegee University Center for Food Production, Processing and Waste Management in Controlled Ecological Life Support Systems (CELSS) is to provide tested information and technologies applicable to bioregenerative food production systems for life support on long-term manned space mission. Specifically, the center is developing information, computer simulated models, methodologies and technology for sweetpotato and peanut biomass production and processing, inclusive of waste management and recycling of these crops selected by NASA for CELSS. The Center is organized into interdisciplinary teams of life scientists and engineers that work together on specific objectives and long-term goals. Integral to the goal of the Center is the development of both basic and applied research information and the training of young scientists and engineers, especially underrepresented minorities that will increase the professional pool in these disciplines and contribute to the advancement of space sciences and exploration.

Source record↗
DOE OSTI2011

RESULTS OF CESIUM MASS TRANSFER TESTING FOR NEXT GENERATION SOLVENT WITH HANFORD WASTE SIMULANT AP-101

SRNL has performed an Extraction, Scrub, Strip (ESS) test using the next generation solvent and AP-101 Hanford Waste simulant. The results indicate that the next generation solvent (MG solvent) has adequate extraction behavior even in the face of a massive excess of potassium. The stripping results indicate poorer behavior, but this may be due to inadequate method detection limits. SRNL recommends further testing using hot tank waste or spiked simulant to provide for better detection limits. Furthermore, strong consideration should be given to performing an actual waste, or spiked waste demonstration using the 2cm contactor bank. The Savannah River Site currently utilizes a solvent extraction technology to selectively remove cesium from tank waste at the Multi-Component Solvent Extraction unit (MCU). This solvent consists of four components: the extractant - BoBCalixC6, a modifier - Cs-7B, a suppressor - trioctylamine, and a diluent, Isopar L{trademark}. This solvent has been used to successfully decontaminate over 2 million gallons of tank waste. However, recent work at Oak Ridge National Laboratory (ORNL), Argonne National Laboratory (ANL), and Savannah River National Laboratory (SRNL) has provided a basis to implement an improved solvent blend. This new solvent blend - referred to as Next Generation Solvent (NGS) - is similar to the current solvent, and also contains four components: the extractant - MAXCalix, a modifier - Cs-7B, a suppressor - LIX-79{trademark} guanidine, and a diluent, Isopar L{trademark}. Testing to date has shown that this 'Next Generation' solvent promises to provide far superior cesium removal efficiencies, and furthermore, is theorized to perform adequately even in waste with high potassium concentrations such that it could be used for processing Hanford wastes. SRNL has performed a cesium mass transfer test in to confirm this behavior, using a simulant designed to simulate Hanford AP-101 waste.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

A cradle to grave decontamination process - 15266

As the rate of decommissioning of nuclear facilities is accelerating there is an ever increasing need for technologies which will decrease dose to operators, reduce overall cost and allow minimization of waste volumes. A new fully integrated decontamination and waste management process is described herein. This method approaches surface decontamination in an integrated fashion in that it deals with the full cycle of the process including the final disposition of the secondary wastes generated during the cleaning cycle. This approach also simplifies decontamination in that it addresses issues observed in previous chemical decontamination applications such as generation of orphan wastes and the requirements for multi-step reagent additions over repeated cycles. The waste management aspect of this system is key to providing a fully integrated process and deals with the issues related to grout formulations in the presence of high waste loadings. (authors)

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

BETO 2021 Peer Review - Waste Carbon Gas Upgrading via Acetogens 2.3.2.106

Waste carbon gas represents a large and diverse set of feedstocks that could be captured and turned into useful products. This includes waste gas emitted from industrial activity, syngas from burned plant biomass or processed municipal waste, and electrochemical reduction of CO2. Currently, carbon gas is being microbially converted to ethanol as a main product. However, ethanol is a lower value product with a limited market size. While these microbes can make other higher value products, there are no commercial processes for generating these other products, leaving a gap in understanding potential implementation for commercialization. Expanding the products microbially produced from waste carbon gas requires several steps before commercial implementation. We are studying the acetogen Clostridium ljungdahlii as a biocatalyst to convert waste carbon gas to the chemical 3-hydroxybutyrate (3HB), a plastic monomer and fuel precursor. For that, we are studying and engineering microbial characteristics for novel 3HB product formation from waste gas streams. This includes metabolic characterization, genetic engineering, gas fermentation scaling, as well as technoeconomic and life cycle analysis.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2011

Environmental evaluation of municipal waste prevention

Highlights: > Influence of prevention on waste management systems, excluding avoided production, is relatively minor. > Influence of prevention on overall supply chain, including avoided production is very significant. > Higher relative benefits of prevention are observed in waste management systems relying mainly on landfills. - Abstract: Waste prevention has been addressed in the literature in terms of the social and behavioural aspects, but very little quantitative assessment exists of the environmental benefits. Our study evaluates the environmental consequences of waste prevention on waste management systems and on the wider society, using life-cycle thinking. The partial prevention of unsolicited mail, beverage packaging and food waste is tested for a 'High-tech' waste management system relying on high energy and material recovery and for a 'Low-tech' waste management system with less recycling and relying on landfilling. Prevention of 13% of the waste mass entering the waste management system generates a reduction of loads and savings in the waste management system for the different impacts categories; 45% net reduction for nutrient enrichment and 12% reduction for global warming potential. When expanding our system and including avoided production incurred by the prevention measures, large savings are observed (15-fold improvement for nutrient enrichment and 2-fold for global warming potential). Prevention of food waste has the highest environmental impact saving. Prevention generates relatively higher overall relative benefit for 'Low-tech' systems depending on landfilling. The paper provides clear evidence of the environmental benefits of waste prevention and has specific relevance in climate change mitigation.

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

Formulation and preparation of Hanford Waste Treatment Plant direct feed low activity waste Effluent Management Facility core simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations 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. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other problems such a recycle stream present. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to formulate and prepare a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations. That simulant can be used in evaporator testing to predict the composition of the effluents from the Effluent Management Facility (EMF) evaporator to aid in planning for their disposition. This document describes the method used to formulate a simulant of this LAW Melter Off-Gas Condensate stream, which, after pH adjustment, is the feed to the evaporator in the EMF.

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

2017 NE-4 Nuclear Technologies Research and Development (NTRD) Achievements Report

Fiscal year (FY) 2017 initiated new and reestablished presidential and departmental nuclear energy priorities for the U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE-NE) that will: • Support the Administration’s priorities to achieve U.S. energy dominance and economic prosperity through U.S. technology leadership • Support DOE-NE mission priorities and goals for the existing nuclear fleet, advanced reactor pipeline and national fuel cycle infrastructure • Focus on early stage research, and effective cooperation with industry, academia, and other government agencies • Propose investments to maintain and improve capabilities and infrastructure to conduct leading-edge science, research, and technology development. DOE-NE’s mission priorities will focus on the existing nuclear fleet, advanced reactor pipeline, and fuel cycle infrastructure. Therefore, the DOE Nuclear Technology Research and Development (NE-4) mission aligned to (1) support the research and development (R&D) of innovative reactor technologies that may offer improved safety, functionality, and affordability and (2) conduct research to reduce long-term technical barriers for advanced nuclear energy systems, including space fission systems. Key activities to execute this mission include: • To achieve the NE-4 mission priorities, two major programs with eight R&D campaigns are engaged in impactful research, development, demonstration, and deployment: Fuel Cycle Technologies (FCT) Program • Fuel Cycle Options Campaign evaluates complex fuel cycle options and provides performance metrics on various systems and scenarios to improve the knowledge of interdependencies between technologies and different systems. • Advanced Fuels Campaign performs science-based R&D on accident tolerant fuels in addition to advanced light-water reactor (LWR) fuel and transmutation fuel. • Joint Fuel Cycle Studies (JFCS) is a collaboration with the Republic of Korea to assess the feasibility and nonproliferation of electrochemical recycling to manage used fuel. • Material Recovery and Waste Form Development (MRWFD) Campaign researches advanced fuel cycle material recovery and waste management capabilities to improve fuel cycle performance with less processing and waste generation. • Material Protection, Control, and Accountability Technologies (MPACT) Program develops tools and capabilities to secure the next generation nuclear materials management and safeguards for nuclear fuel cycles. Advanced Reactor Technologies (ART) Program • Fast Reactors Campaign targets benefits in either resource utilization or waste management and possesses favorable features for small reactor applications and plutonium management. • Gas-Cooled Reactors Campaign seeks to raise the technological readiness of high temperature gas-cooled reactors (HTGRs) and other reactor concepts utilizing coated particle fuel, graphite, or helium coolant by performing high risk, enabling research and development that commercial developers are unable or unwilling to undertake in the current energy market. • ART Energy Conversion (EC) Project provides solutions to convert heat from an advanced reactor to useful products that support commercial application of the reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1986

Controlled Ecological Life Support Systems (CELSS) physiochemical waste management systems evaluation

Parametric data for six waste management subsystems considered for use on the Space Station are compared, i.e.: (1) dry incineration; (2) wet oxidation; (3) supercritical water oxidation; (4) vapor compression distillation; (5) thermoelectric integrated membrane evaporation system; and (6) vapor phase catalytic ammonia removal. The parameters selected for comparison are on-orbit weight and volume, resupply and return to Earth logistics, power consumption, and heat rejection. Trades studies are performed on subsystem parameters derived from the most recent literature. The Boeing Engineering Trade Study (BETS), an environmental control and life support system (ECLSS) trade study computer program developed by Boeing Aerospace Company, is used to properly size the subsystems under study. The six waste treatment subsystems modeled in this program are sized to process the wastes for a 90-day Space Station mission with an 8-person crew, and an emergency supply period of 28 days. The resulting subsystem parameters are compared not only on an individual subsystem level but also as part of an integrated ECLSS.

Oleson, M.↗
DOE OSTI2017

Fabrication and Properties of Advanced Ceramic Technetium Waste Forms - 17311

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Evaluating the Impacts of Reactive Gas (Carbon Dioxide and Oxygen) Ingress on Leaching from Cementitious Waste Forms - 17482

Cementitious materials are currently being used as the waste forms for low-activity waste disposal (Saltstone) at the Savannah River Site. In addition, cementitious materials have been selected as the waste form (Cast Stone) for secondary wastes and are being investigated for use for supplemental low-activity waste treatment (Cast Stone) at the Hanford Site. These materials are subjected to numerous aging and degradation phenomena in the disposal environment that potentially impact their long-term durability and performance. One important phenomenon is the ingress of reactive gases (i.e., carbon dioxide and oxygen) that can react with and change the chemistry and pore structure and thus mass transport properties of the waste forms. The reaction of carbon dioxide (i.e., carbonation) reduces the pH of the waste form pore solution impacting the solubility of constituents of potential concern (COPCs). In Cast Stone and Saltstone, the reaction of oxygen (i.e., oxidation) impacts the solubility of redox-sensitive COPCs, notably the technetium-99 (Tc-99) and chromium. The objective of this research is to experimentally characterize the carbonation and oxidation impacts on chemistry and leaching behavior in cementitious waste forms using experimental techniques including United States Environmental Protection Agency (U.S. EPA) Method 1313 and U.S. EPA Method 1315. Results of experimental characterization will be used to parameterize reactive mass transport models to project leaching under a range of field scenarios in future applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Solidification of Low and Intermediate Level Waste - A Comparison between Different Methods - 19584

The treatment of radioactive materials results in products that have to be suitable for a period of interim storage as well as for disposal in a repository. Considering available options for treatment of radioactive waste, future disposal also has to be taken into account. Up to now, only cementation and encapsulation in geopolymer (only to limited extent) are relevant treatment methods for the solidification of German waste for disposal in the repository Grube Konrad. However, a synopsis of other solidification technologies is considered to be helpful in order to be prepared for future developments in the fields of radioactive waste treatment. Treatment methods addressed in this paper are: - Cementation; - Encapsulation into geopolymer matrix; - Vitrification; - SYNROC; - Geomelting. The criteria for evaluation of the treatment methods are as follows: - Requirements concerning waste properties; - Safety of treatment facilities; - Fulfillment of waste acceptance criteria for repository Grube Konrad; - Fulfillment of requirements for near surface disposal; - Suitability for long-term interim storage. (authors)

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

Regulatory Framework for Salt Waste Disposal and Tank Closure at the Savannah River Site - 13663

The end of the Cold War has left a legacy of approximately 37 million gallons of radioactive waste in the aging waste tanks at the Department of Energy's Savannah River Site (SRS). A robust program is in place to remove waste from these tanks, treat the waste to separate into a relatively small volume of high-level waste and a large volume of low-level waste, and to actively dispose of the low-level waste on-site and close the waste tanks and associated ancillary structures. To support performance-based, risk-informed decision making and to ensure compliance with all regulatory requirements, the U.S. Department of Energy (DOE) and its current and past contractors have worked closely with the South Carolina Department of Health and Environmental Control (SCDHEC), the U.S. Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) to develop and implement a framework for on-site low-level waste disposal and closure of the SRS waste tanks. The Atomic Energy Act of 1954, as amended, provides DOE the authority to manage defense-related radioactive waste. DOE Order 435.1 and its associated manual and guidance documents detail this radioactive waste management process. The DOE also has a requirement to consult with the NRC in determining that waste that formerly was classified as high-level waste can be safely managed as either low-level waste or transuranic waste. Once DOE makes a determination, NRC then has a responsibility to monitor DOE's actions in coordination with SCDHEC to ensure compliance with the Title 10 Code of Federal Regulations Part 61 (10CFR61), Subpart C performance objectives. The management of hazardous waste substances or components at SRS is regulated by SCDHEC and the EPA. The foundation for the interactions between DOE, SCDHEC and EPA is the SRS Federal Facility Agreement (FFA). Managing this array of requirements and successfully interacting with regulators, consultants and stakeholders is a challenging task but ensures thorough and thoughtful processes for disposing of the SRS low-level waste and the closure of the tank farm facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗