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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

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DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) as a general strategy and associated Waste Treatment Baseline Study (WTBS). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to both a nominal case study and near-term target reprocessing flowsheets. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include: • Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers (HICs). • Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. • Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. • Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. • An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased crystalline synthetic rock (Synroc). • An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or Synroc. • Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative with high promise the purification and reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations. • Electrochemical process salt HLW which is incorporated into a glass bonded Sodalite waste form known as the ceramic waste form (CWF). • Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from the reference mass balances are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Vitrification of Hanford Tank Waste 241-AP-107 in a Continuous Laboratory-Scale Melter

Low-activity waste (LAW) stored in underground tanks on the Hanford Site in Washington State is planned to be filtered for solids removal and processed through ion exchange columns for cesium removal. These pretreatment steps will allow the waste to be transferred to the Hanford Tank Waste Treatment and Immobilization Plant’s LAW Facility for immobilization into glass. The liquid waste will be combined with glass-forming chemicals (GFCs) to form a waste feed slurry that can be fed to electric melters for vitrification. The process of continuously converting the aqueous feed slurry into a melt is dynamic and includes multiple reactions, degassing, and dissolution processes that depend on heat from the melt below. In this conversion process, waste components are partitioned into one of two streams: glass and off-gas. Washington River Protection Solutions has requested processing information and chemical information associated with these waste products for actual waste from Hanford tank 241-AP-107 (referred to herein as AP-107). To acquire this type of information, a small-scale melter system was designed that would not require high volumes of input waste or the large resource commitment of a full-scale melter system, while also providing dynamic information that would be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM) has been designed to operate with a continuous feeding process while periodically pouring glass product and collecting off-gas. The CLSM vessel has been sized to collect the relevant process and chemical information from obtainable volumes of AP-107 waste samples. A total 8.6 L of actual AP-107 tank waste were received after filtering for solids removal and ion exchange for cesium removal. This volume of AP-107 waste was mixed with GFCs to form an estimated 11.8 L of melter feed slurry. A mass of 7.01 kg of glass product were poured from the CLSM vessel during 10.07 hours of charging the AP-107 melter feed, indicating that about 15.0 kg of melter feed was vitrified into glass. The CLSM system was also designed with the capability to fully divert the flow of off-gas produced in the CLSM vessel to a sampling line that could capture the volatile species of interest, such as technetium-99 ( 99 Tc). This novel sampling system avoided the difficulties of slipstream sampling and could be activated once the feeding reached a steady state within the CLSM vessel. Off-gas product samples captured via high-efficiency particulate air (HEPA) filters using this method, as well as selected glass product samples were sent to an analytical lab for chemical analysis.

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DOE OSTI2017

Challenges to Collection of High Quality Samples of Saltstone - 17468

The Savannah River Site (SRS) Saltstone Disposal Facility (SDF) is used to dispose of low-level radioactive liquid waste in the form of grout. The SDF is comprised of a series of concrete saltstone disposal units (SDUs) that receive the grout. The grout solidifies into a cementitious waste form known as saltstone. Properties of saltstone that influence contaminant release have been measured on simulant material but not on samples of emplaced saltstone because samples of emplaced radioactive saltstone are difficult to obtain, transport, and store in a manner that protects the properties of interest. Savannah River Remediation (SRR) initiated a project to obtain samples of emplaced saltstone such that the two critical properties, hydraulic conductivity and reduction capacity, are negligibly impacted by the sample collection effort. The sample collection project was established to manage sample collection efforts beginning with sample collection mock-up activities through field execution. The project focused on three phases: mock-up in a clean environment, development of equipment needed to maintain an inert environment during sample transport and laboratory storage, and coordination of Facility operations and support resources and schedules to support sample collection. A mock-up facility was constructed at SRS to mimic the SDU. Core drill equipment was tested, procedures developed, and operator training conducted using the mockup facility. Concurrent with mock-up activities, inert transportation and storage equipment was designed, manufactured, tested, and specific operating instruction included in the sample collection procedure. Inerting equipment minimized oxidation from the atmosphere as the samples were removed from the disposal unit, transported, and stored in the laboratory awaiting analysis. Coordination with the Saltstone Facility began during construction of the mock-up platform to collect current as-built information for platform construction. Baseline data was collected from the disposal unit vapor space to assess the potential for personnel exposure from chemical hazards associated with trapped vapors. Based on this data, mitigating actions were incorporated into the project scope. Finally, project and facility schedules were integrated to assure that facility non-processing windows, support personnel, equipment fabrication, and field preparation activities were in alignment and supportive of the overall sample collection objectives. During May 2014, 485 linear cm of approximately 5 cm diameter radioactive core material was successfully collected from three access ports in SDU Cell 2A, inerted, transported to the laboratory, and stored in an inert environment for analysis. (authors)

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DOE OSTI2017

Mercury in aqueous tank waste at the Savannah River Site: Facts, forms, and impacts

Over the past two years, there has been an intense effort to understand the chemistry of mercury across the Savannah River Site’s high-level liquid waste system to determine the impacts of various mercury species. This effort started after high concentrations of mercury were measured in the leachates from a toxicity characteristic leaching procedure (TCLP) test on the low-level cementitious waste form produced in the Savannah River Saltstone facility. Speciation showed the dominant form of leached mercury to be the methylmercury cation. Neither the source of the methylmercury nor its concentration in the Saltstone feed was well established at the time of the testing. Finally, this assessment of mercury was necessary to inform points in the process operations that may be subject to new separation technologies for the removal of mercury.

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DOE OSTI2018

Vitrification of Hanford Tank Waste 241-AP-105 in a Continuous Laboratory-Scale Melter

Low-activity waste (LAW) stored in underground tanks on the Hanford site in Washington State is planned to be filtered for solids removal and processed through ion exchange columns for cesium removal at the Low-Activity Waste Pretreatment System facility. These pretreatment steps will allow the waste to be transferred to the Hanford Tank Waste Treatment and Immobilization Plant LAW facility for immobilization into glass. The liquid waste will be combined with glass-forming chemicals (GFCs) to form a waste feed slurry that can be fed to electric melters for vitrification. The process of continuously converting the aqueous feed slurry into a melt is dynamic and includes multiple reactions, degassing, and dissolution processes that depend on heat from the melt below. In this conversion process, waste components are partitioned into one of two streams: glass and off-gas. Processing information and chemical information associated with these waste product streams has been requested by Washington River Protection Solutions for actual tank waste from tank 241-AP-105 (referred to as AP-105). To acquire this type of information, a small-scale melter system was desired that would not require high volumes of input waste or the large resource commitment of a full-scale melter system, while also providing dynamic information that would be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM), has been designed to operate with a continuous feeding process, while periodically pouring glass product and collecting off-gas. The CLSM vessel has been sized to collect the relevant process and chemical information from obtainable volumes of AP-105 waste samples. Two CLSMs were constructed: one in a non-radioactive environment for processing waste simulants and another in a fume hood capable of handling radioactive material for processing actual Hanford tank waste. This dual-system setup allowed for comparison between simulated and real tank waste with essentially identical test equipment. AP-105 waste simulant was first tested in the non-radioactive CLSM to determine the processability and desired operating window for the actual tank waste. Rhenium (Re) was added to the simulant to act as a surrogate for technetium-99 ( 99 Tc), a key waste component found in the actual tank waste. The partitioning of Re into the glass and off-gas products during the continuous feeding of simulant was determined, which allowed the Re retention, the mass flow rate of Re out of the CLSM via the glass product with respect to the mass flow rate of Re into the CLSM via the feed slurry, to be calculated. Key processing results and the average Re retention, from the CLSM test run with AP-105 simulant feed slurry are shown in Table ES.1.

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DOE OSTI2017

EPRI Research Results: International Nuclear Power Plant Waste Classification - 17444

Different countries around the globe use varying schemes for the classification of radioactive waste. Many countries develop country specific waste classification schemes based on the International Atomic Energy Agency (IAEA) waste classification framework. The United States of America use the Class A, B, and C waste classification. Additionally, some countries use concentration limits to classify wastes while others use dose rate limits or a combination of concentration and dose rate limits. General comparisons have been made between the IAEA and the United States classification schemes, but a clear and well researched comparison of the international schemes was not evident in the available research data sets. In 2013 Electric Power Research Institute (EPRI) initiated the Comparison of Global Low and Intermediate Level Waste Management Methods project. The objective of the project was to understand how six countries managed their low and intermediate level wastes from generation to disposal. A major task of this project was to conduct a comparison of the international waste classification schemes. EPRI developed a set of reference waste volumes that are representative of nuclear power plant low-level radioactive waste in form and radioactivity concentrations (e.g. dry active waste, high activity resin, low activity resin, high activity filters and low activity filters). These waste volumes were then classified using the waste classification systems in place in several countries around the world. In 2016, EPRI presented a paper providing a status update on this research. This 2017 paper provides the final results of that research, show where comparisons and correlations can be made, and provide a discussion on how this information might inform international radioactive research and development and benchmarking of radioactive waste management strategies, operational experiences, best practices, and technology implementation. (authors)

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DOE OSTI2016

EPRI Project Update on International Nuclear Power Plant Waste Classification - 16224

Countries around the globe use varying schemes for the characterization of radioactive waste. Many countries use adaptations of the International Atomic Energy Agency (IAEA) waste classification scheme. Other countries use the Class A, B, and C waste classification scheme that was developed in the United States. General comparisons have been made between the IAEA and the United States (U.S.) classification schemes, but a clear and well researched comparison of the two schemes was not evident in the available research data sets. The Electric Power Research Institute (EPRI) Radiation Safety (RS) Program has conducted a technical evaluation that compares these two waste characterization schemes. EPRI developed a set of reference wastes that are representative of nuclear power plant low-level radioactive waste in form and radioactivity concentrations (e.g. dry active waste, high activity resin, low activity resin, high activity filters and low activity filters). These reference wastes were then classified using the waste classification systems in place in several countries around the world. This paper and presentation will provide a summary of that research, show where strong comparisons and correlations can be made, and provide a discussion on how this information might inform the harmonization of global waste classification standards. (authors)

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DOE OSTI2019

Impact of Crystalline Silicotitanate Particle Size on Cesium Removal Efficiency

Washington River Protection Solutions is working on a Tank-Side Cesium Removal (TSCR) system to remove Cs from Hanford tank waste supernate. Removal of 137 Cs from tank waste supernate by TSCR results in a low-activity waste (LAW) form that can be fed directly to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW vitrification facility, where the supernate can be vitrified into a glass form. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIV TM R9140-B), was selected for use as the ion exchange media to remove Cs at TSCR. The particle size used in testing is expected to impact the kinetic performance of the ion exchange media. Therefore, two single-column systems were tested with CST Lot 2002009604. One column contained <25 mesh sieve cut fraction and the other column contained <35 mesh sieve cut fraction. Flowrate through both test systems was adjusted to match the CST contact time expected for the full-scale operation, i.e., matched bed volumes per hour (BV/h) at 1.83 BV/h. The feed was processed downflow through the columns until ~1000 BVs of feed were processed. The waste acceptance criteria (WAC) limit for the WTP LAW vitrification facility is <3.18E-5 Ci 137 Cs per mole Na. For a tank waste containing 156 µCi/mL 137 Cs and 5.6 M Na, up to 0.114% of the influent 137 Cs concentration may be delivered to the WTP (based on 137 Cs content in AP-107 tank waste); this required a Cs decontamination factor of 876. The WAC limit was used as one of the relevant process parameters, along with the 50% breakthrough point, Cs loading per gram of CST, and the mass transfer zone.

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DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Comparison of High Level Waste Glass Feeds Containing Frit and Glass Forming Chemicals - 16154

This study compares the impact of frit and glass forming-chemicals on high level waste feed melting behavior. Feeds formulated to yield high-boron (lower waste loading) and low-boron (higher waste loading) glass compositions were prepared by blending a simulated high-level waste with glass-forming additives either in the form of chemicals or pre-melted as glass frit. The feeds were pressed into cylindrical disc-shaped pellets using a die tool and a hydraulic press. Each pellet was heat-treated in a high-temperature furnace equipped with a window to view the pellet's response to increasing temperature, usually up to 1100 deg. C, until the pellet turned into a molten glass droplet. A camera mounted in front of the viewing window was used to take high-quality pictures at predetermined intervals. Images were analyzed using a program that calculated the pellet profile area and volume as functions of temperature. The analyses are used to determine characteristic temperatures and associated volume expansions including foam onset and collapse. (authors)

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DOE OSTI2016

Technetium Management for Hanford Tank Waste Processing and Disposition - 16399

Technetium-99 (Tc) generated from the fission of {sup 235}U and {sup 239}Pu in high yields is one of the most difficult contaminants to be addressed at the U.S. Department of Energy (DOE) Hanford Site. Among radioactive constituents in the tank waste, Tc presents a unique challenge in that it is radio-toxic (β = 292 keV), has a long half-life (T{sub 1/2} = 2.11x10{sup 5} y), and exists predominantly in the liquid fraction of the alkaline tank waste, generally in the anionic form of pertechnetate TcO{sub 4}{sup -}, which is highly volatile at low-activity waste vitrification melter temperatures and mobile in the subsurface environment. This makes immobilization of Tc into high-durability waste forms a critical technical challenge for nuclear waste management. The complex behavior of Tc under storage, treatment, and immobilization conditions significantly affects its management options. The overall objective of the Technetium Management program is to provide the DOE Office of Environmental Management with practical solutions to Tc needs encountered in the processing of Hanford tank waste. To reduce the technical uncertainty associated with the long-term environmental impact of Tc, a systems-based approach is needed to address Tc management needs by evaluating and developing science and engineering options for treatment, immobilization, and disposition of Tc contained in Hanford tank wastes. This paper describes major elements of Technetium Management program, including: - Identifying the chemical forms of Tc in tank waste through analytical speciation of low-valent Tc(I) species in alkaline solutions. - Performing proof-of-concept demonstration for detection of low-valent Tc(I) by fluorescence spectroscopy. - Identifying alternative non-glass Tc immobilization methods. (authors)

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DOE OSTI2012

La Hague Legacy Waste Recovery Program: Scope, Progress and Issues -12080

A significant inventory of process waste of varying natures and quantities has been generated during the thirty years of operation of UP2 400 facility on the site of La Hague, France. The retrieval, packaging and final storage of such an inventory has never been achieved before in France and thus requires the design and qualification of new processes, equipment, and waste packages. Following AREVA strategic decisions and French safety authority requirements, the legacy waste program has begun around the year 2000 and is scheduled to be completed around the year 2025. It is under the responsibility of AREVA Site Value Development Project teams. For each category of waste to be recovered, AREVA teams conducted detailed investigations, defined recovery modes, treatment processes, as well as final waste package forms, which they subsequently submitted to French safety and waste management authorities. A Task force initiative was subsequently launched to optimize the program cost and scenario, and lead to an optimization of about 15% of the entire program. The qualification of processes and waste packages required a significant amount of research and development which is now well under way for processes, and scheduled to be completed in 2015. Preparation work has begun on several installations to clear space for the construction of future retrieval facilities, scheduled to begin in the coming three years. La Hague Legacy waste retrieval program represents a significant challenge in the sense that it covers a significant variety and quantity of waste needing recovery and reconditioning, with tight financial objectives and a binding recovery schedule. During the past five years, AREVA SVD successfully conducted design, research, development, and qualification activities which lead to the definition of qualified processes and waste packages for each retrieval program. Preparation work and supplier consultations are now on-going, in order to meet our objectives of beginning retrieval operations in compliance with our commitments to the safety authorities, in 2015 and 2016. (author)

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DOE OSTI2016

Analyses of Integrated Waste Treatment Unit (IWTU) TPR-8023 (1&2) Samples Including Simulant, Bed Products, and Wall Scale Formed during Fluidized Bed Steam Reforming of Sodium Bearing Waste into a Carbonate Form

This report documents Savannah River National Laboratory (SRNL) analyses of the products from the various vessels in the Integrated Waste Treatment Unit (IWTU) at the Idaho Nuclear Technology and Engineering Center (INTEC) during non-radioactive simulant processing of campaign TPR-8023 (1&2) in November/December 2015. The analyses focus on the identification of unwanted materials deposited in the IWTU during that campaign. The IWTU is designed to stabilize the acidic low-activity waste known as Sodium Bearing Waste (SBW) through a Fluidized Bed Steam Reformer (FBSR) system. The THOR® FBSR mineralizing technology uses dual reformers to pyrolyze organics in the presence of a fluidization media of steam. Organics not pyrolyzed in the Denitration Mineralization Reformer (DMR) and excess hydrogen are oxidized in the second reformer known as the Carbon Reduction Reformer (CRR).

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

Immobilization of organic radioactive and non-radioactive liquid waste in a composite matrix

A method for immobilizing liquid radioactive waste is provided, the method having the steps of mixing waste with polymer to form a non-liquid waste; contacting the non-liquid waste with a solidifying agent to create a mixture, heating the mixture to cause the polymer, waste, and filler to irreversibly bind in a solid phase, and compressing the solid phase into a monolith. The invention also provides a method for immobilizing liquid radioactive waste containing tritium, the method having the steps of mixing liquid waste with polymer to convert the liquid waste to a non-liquid waste, contacting the non-liquid waste with a solidifying agent to create a mixture, heating the mixture to form homogeneous, chemically stable solid phase, and compressing the chemically stable solid phase into a final waste form, wherein the polymer comprises approximately a 9:1 weight ratio mixture of styrene block co-polymers and cross linked co-polymers of acrylamides.

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

Scoping Tests of Technetium and Iodine Removal from Tank Waste Using SuperLig® 639 Resin

The primary chemical form of 99 Tc found in Hanford Low Activity Waste (LAW) is pertechnetate anion (TcO4 -), which is highly soluble in water, and is mobile if released to the environment. Pertechnetate will not be removed from the aqueous waste in the Hanford waste treatment plant, and the primary disposition path is immobilization in the LAW glass waste form, which will be disposed in the Integrated Disposal Facility (IDF). Due to the soluble properties of pertechnetate, and the potential for impact to the Performance Assessment (PA), effective management of 99 Tc is important to the overall success of the River Protection Project mission. Washington River Protection Solutions (WRPS) is developing some conceptual flow-sheets for LAW treatment and disposal that could benefit from technetium removal. While 99 Tc is the primary radionuclide of interest, 129 I also contributes to the calculated future dose of disposed LAW, and it would be of interest to examine if removal is possible.

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

Estimating Residual Solids Volume In Underground Storage Tanks

The Savannah River Site liquid waste system consists of multiple facilities to safely receive and store legacy radioactive waste, treat, and permanently dispose waste. The large underground storage tanks and associated equipment, known as the 'tank farms', include a complex interconnected transfer system which includes underground transfer pipelines and ancillary equipment to direct the flow of waste. The waste in the tanks is present in three forms: supernatant, sludge, and salt. The supernatant is a multi-component aqueous mixture, while sludge is a gel-like substance which consists of insoluble solids and entrapped supernatant. The waste from these tanks is retrieved and treated as sludge or salt. The high level (radioactive) fraction of the waste is vitrified into a glass waste form, while the low-level waste is immobilized in a cementitious grout waste form called saltstone. Once the waste is retrieved and processed, the tanks are closed via removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. The comprehensive liquid waste disposition system, currently managed by Savannah River Remediation, consists of 1) safe storage and retrieval of the waste as it is prepared for permanent disposition; (2) definition of the waste processing techniques utilized to separate the high-level waste fraction/low-level waste fraction; (3) disposition of LLW in saltstone; (4) disposition of the HLW in glass; and (5) closure state of the facilities, including tanks. This paper focuses on determining the effectiveness of waste removal campaigns through monitoring the volume of residual solids in the waste tanks. Volume estimates of the residual solids are performed by creating a map of the residual solids on the waste tank bottom using video and still digital images. The map is then used to calculate the volume of solids remaining in the waste tank. The ability to accurately determine a volume is a function of the quantity and quality of the waste tank images. Currently, mapping is performed remotely with closed circuit video cameras and still photograph cameras due to the hazardous environment. There are two methods that can be used to create a solids volume map. These methods are: liquid transfer mapping / post transfer mapping and final residual solids mapping. The task is performed during a transfer because the liquid level (which is a known value determined by a level measurement device) is used as a landmark to indicate solids accumulation heights. The post transfer method is primarily utilized after the majority of waste has been removed. This method relies on video and still digital images of the waste tank after the liquid transfer is complete to obtain the relative height of solids across a waste tank in relation to known and usable landmarks within the waste tank (cooling coils, column base plates, etc.). In order to accurately monitor solids over time across various cleaning campaigns, and provide a technical basis to support final waste tank closure, a consistent methodology for volume determination has been developed and implemented at SRS.

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