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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Early Regulatory Engagement for Successful Site Remediation: the UK Experience - 13173

The Office for Nuclear Regulation (ONR) is an independent safety, security and transport regulator of the UK nuclear industry. ONR regulates all civil nuclear reactor power stations, fuel manufacture, enrichment, spent fuel reprocessing, most defence sites and installations that store and process legacy spent fuel and radioactive waste. The responsibility for funding and strategic direction of decommissioning and radioactive waste management of state owned legacy sites has rested solely with the Nuclear Decommissioning Authority (NDA) since 2005. A key component of NDA's mandate was to encourage new strategic approaches and innovation to dealing with the UK's waste legacy and which deliver value-for-money to the UK taxpayer. ONR, as an agency of the Health and Safety Executive, is entirely independent of NDA and regulates all prescribed activities on NDA's sites. NDA's competition of site management and closure contracts has attracted significant international interest and the formation of consortia comprised of major British, US, French and Swedish organizations bidding for those contracts. The prominence of US organizations in each of those consortia reflects the scale and breadth of existing waste management and D and D projects in the US. This paper will articulate, in broad terms, the challenges faced by international organizations seeking to employ 'off-the-shelf' technology and D and D techniques, successfully employed elsewhere, into the UK regulatory context. The predominantly 'goal-setting' regulatory framework in the UK does not generally prescribe a minimum standard to which a licensee must adhere. The legal onus on licensees in the UK is to demonstrate, whatever technology is selected, that in its applications, risks are reduced 'So Far As Is Reasonably Practicable' or 'SFAIRP'. By the nature of its role, ONR adopts a conservative approach to regulation; however ONR also recognises that in the decommissioning (and ultimately the site closure) domain, it is often necessary to consider and support novel approaches to achieve the nationally desired end-state. Crucial to successful and compliant operation in this regulatory environment is early and sustained engagement of the contractor with the regulator. There must be a 'no-surprises' culture to engender regulatory confidence early in a project. The paper considers some of the challenges facing international prime and lower tier contractors when undertaking D and D contracts in the UK, and emphasizes the importance of constructive and transparent dialogue with all regulators to sustain confidence at all stages of a major decommissioning project. The paper will also articulate ONR's strategy to increase collaboration with the US Department of Energy in light of increasing UK-US synergy in the area of waste management and to benchmark respective regulatory approaches. (authors)

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

Investigating the Impacts of Direct Dissolution Conditions on the Radiolytic Longevity of Butyramide Extractants

Removing the nitric acid (HNO3) dissolution step in used nuclear fuel (UNF) reprocessing would reduce the volume of radioactive waste streams generated, thereby, improving process efficiency. A promising strategy for this is the direct dissolution of UNF that has been pretreated by voloxidation into an organic solvent composed of specialized extractants and diluent. However, removal of the aqueous HNO3 phase from the envisioned reprocessing system has the potential to drastically change the suite of radiation-induced processes occurring, and thus, alter the longevity of proposed reagents. Furthermore, the impacts of fission product and transuranic metal ion complexation on the aforementioned radiation-induced processes is poorly understood, and yet can cause significant changes in radiolytic longevity. To bridge these knowledge gaps and support the continued development of direct dissolution strategies, we present an investigation into the impacts of direct dissolution conditions on the gamma radiation-induced degradation of N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) ligands—candidate replacements for tributyl phosphate—in pre-equilibrated n-dodecane solvent in the presence and absence of envisioned loading amounts of uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Investigating the impacts of used nuclear fuel direct dissolution on the radiolytic longevity of solvent and butyramide extractants

Removing the nitric acid (HNO3) dissolution step in used nuclear fuel (UNF) reprocessing would reduce the volume of radioactive waste streams generated, thereby, improving process efficiency. A promising strategy for this is the direct dissolution of UNF that has been pretreated by voloxidation into an organic solvent composed of specialized extractants and diluent. However, removal of the aqueous HNO3 phase from the envisioned reprocessing system has the potential to drastically change the suite of radiation-induced processes occurring, and thus, alter the longevity of proposed reagents. Furthermore, the impacts of fission product and transuranic metal ion complexation on the aforementioned radiation-induced processes is poorly understood, and yet can cause significant changes in radiolytic longevity. To bridge these knowledge gaps and support the continued development of direct dissolution strategies, we present an investigation into the impacts of direct dissolution conditions on the gamma radiation-induced degradation of N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) ligands—candidate replacements for tributyl phosphate—in pre-equilibrated n-dodecane solvent in the presence and absence of envisioned loading amounts of uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2012

Initial Evaluation of Processing Methods for an Epsilon Metal Waste Form

During irradiation of nuclear fuel in a reactor, the five metals, Mo, Pd, Rh, Ru, and Tc, migrate to the fuel grain boundaries and form small metal particles of an alloy known as epsilon metal (ε-metal). When the fuel is dissolved in a reprocessing plant, these metal particles remain behind with a residue - the undissolved solids (UDS). Some of these same metals that comprise this alloy that have not formed the alloy are dissolved into the aqueous stream. These metals limit the waste loading for a borosilicate glass that is being developed for the reprocessing wastes. Epsilon metal is being developed as a waste form for the noble metals from a number of waste streams in the aqueous reprocessing of used nuclear fuel (UNF) - (1) the ε-metal from the UDS, (2) soluble Tc (ion-exchanged), and (3) soluble noble metals (TRUEX raffinate). Separate immobilization of these metals has benefits other than allowing an increase in the glass waste loading. These materials are quite resistant to dissolution (corrosion) as evidenced by the fact that they survive the chemically aggressive conditions in the fuel dissolver. Remnants of ε-metal particles have survived in the geologically natural reactors found in Gabon, Africa, indicating that they have sufficient durability to survive for ≈ 2.5 billion years in a reducing geologic environment. Additionally, the ε-metal can be made without additives and incorporate sufficient foreign material (oxides) that are also present in the UDS. Although ε-metal is found in fuel and Gabon as small particles (≈10 µm in diameter) and has survived intact, an ideal waste form is one in which the surface area is minimized. Therefore, the main effort in developing ε-metal as a waste form is to develop a process to consolidate the particles into a monolith. Individually, these metals have high melting points (2617°C for Mo to 1552°C for Pd) and the alloy is expected to have a high melting point as well, perhaps exceeding 1500°C. The purpose of the work reported here is to find a potential commercial process with which ε-metal plus other components of UDS can be consolidated into a solid with minimum surface area and high strength Here, we report the results from the preliminary evaluation of spark-plasma sintering (SPS), hot-isostatic pressing (HIP), and microwave sintering (MS). Since bulk ε-metal is not available and companies could not handle radioactive materials, we prepared mixtures of the five individual metal powders (Mo, Ru, Rh, Pd, and Re) and baddeleyite (ZrO 2 ) to send the vendors of SPS, HIP, and MS. The processed samples were then evaluated at the Pacific Northwest National Laboratory (PNNL) for bulk density and phase assemblage with X-ray diffraction (XRD) and phase composition with scanning electron microscopy (SEM). Physical strength was evaluated qualitatively. Results of these scoping tests showed that fully dense cermet (ceramic-metal composite) materials with up to 35 mass% of ZrO 2 were produced with SPS and HIP. Bulk density of the SPS samples ranged from 87 to 98% of theoretical density, while HIP samples ranged from 96 to 100% of theoretical density. Microwave sintered samples containing ZrO 2 had low densities of 55 to 60% of theoretical density. Structurally, the cermet samples showed that the individual metals alloyed in to ε-phase - hexagonal-close-packed (HCP) alloy (4-95 mass %), the α-phase - face-centered-cubic (FCC) alloy structure (3-86 mass %), while ZrO 2 remained in the monoclinic structure of baddeleyite. Elementally, the samples appeared to have nearly uniform composition, but with some areas rich in Mo and Re, the two components with the highest melting points. The homogeneity in distribution of the elements in the alloy is significantly improved in the presence of ZrO 2 . However, ZrO 2 does not appear to react with the alloy, nor was Zr found in the alloy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Radiological mapping of facilities under dismantling process by digital autoradiography - 15230

because safety standards have changed. In that context, radioactive waste management is a challenging task faced by existing nuclear power countries. Therefore, characterization is essential to classify wastes and to reduce costs. In many facilities, after the removal of glove boxes, fume cupboards and so on, the remaining materials on the floor, walls, and ceiling must be characterized. Gamma emitters can be detected remotely by commercial devices. However, analytical techniques to detect alpha and beta emitters in situ are generally not well suited for large area investigations. The Digital Autoradiography (DA) technique, initially dedicated to research in biology, was re-routed for final state characterization before facility demolition. It is based on the exposure of radio-sensitive phosphor screens on the materials to characterize. Phosphor screens, after a scanning process, provide a picture of the radioactivity, and are reusable several tens of times. This analytical technique is sensitive to all types of radioactivity, requires neither operators nor electricity during acquisition process, and produces only very few wastes. DA has already been tested in different facilities to obtain semi-quantitative values of C-14 and H-3 activities. Large amount of data gathered to cover the whole surface to be characterized. A Geographic Information System (GIS) associated with geostatistical method has been used to provide Radiological Mapping. This method to interpret Digital Autoradiography results has been proved very efficient to prepare remediation project. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2011

DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIGUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. A solvent extraction system for removal of cesium from alkaline solutions was developed utilizing a novel solvent invented at the Oak Ridge National Laboratory (ORNL). This solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix. A modifier is added to the solvent to enhance the extraction power of the calixarene and to prevent the formation of a third phase. An additional additive is used to improve stripping performance and to mitigate the effects of any surfactants present in the feed stream. The process that deploys this solvent system is known as Caustic Side Solvent Extraction (CSSX). The solvent system has been deployed at the Savannah River Site (SRS) in the Modular CSSX Unit (MCU) since 2008.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

TEX 2.0: DOE-EM NCS Experimental Needs [Slides]

This presentation touches on the needs presented from recent work with Hanford Tank Farms (HTF) and Savannah River Site (SRS) Liquid Waste Process (LWP). The Meetings were held with other sites by presenting information about TEX 2.0 and having discussions on what the current challenges were for criticality safety validation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2014

Scientific Solutions to Nuclear Waste Environmental Challenges

The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount of time it takes for one-half of the material to undergo radioactive decay.) In general, the ideal material would need to be durable for approximately 10 half-lives to allow the activity to decay to negligible levels. However, the potential health effects of each radionuclide vary depending on what type of radiation is emitted, the energy of that emission, and the susceptibility for the human body to accumulate and concentrate that particular element. Consequently, actual standards tend to be based on limiting the dose (energy deposited per unit mass) that is introduced into the environment. The Environmental Protection Agency (EPA) has the responsibility to establish standards for nuclear waste disposal to protect the health and safety of the public. For example, the Energy Policy Act of 1992 directed the EPA to establish radiation protection standards for the Yucca Mountain geologic repository for nuclear wastes. The standards for Yucca Mountain were promulgated in 2008, and limit the dose to 15 millirem per year for the first 10,000 years, and 100 milirem per year between 10,000 years and 1 million years (40 CFR Part 197; http://www.epa.gov/radiation/yucca/2008factsheet.html). So, the challenge is two-fold: (1) develop a material (a waste form) that is capable of immobilizing the waste over geologic time scales, and (2) develop a process to convert the radioactive sludge in the tanks into this durable waste form material. Glass: Hard, durable, inert, and with infinite chemical versatility Molten glass is a powerful solvent liquid, which can be designed to dissolve almost anything. When solidified, it can be one of the most chemically inert substances known to man. Nature's most famous analogue to glass is obsidian, a vitreous product of volcanic activity; formations over 17 million years old have been found. Archaeologists have found man-made glass specimens that are five thousand years old.

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

Waste Information Management System-2012 - 12114

The Waste Information Management System (WIMS) -2012 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 40 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and inter-modal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. WIMS continues to successfully accomplish the goals and objectives set forth by DOE for this project. It has replaced the historic process of each DOE site gathering, organizing, and reporting their waste forecast information utilizing different databases and display technologies. In addition, WIMS meets DOE's objective to have the complex-wide waste forecast and transportation information available to all stakeholders and the public in one easy-to-navigate system. The enhancements to WIMS made since its initial deployment include the addition of new DOE sites and facilities, an updated waste and transportation information, and the ability to easily display and print customized waste forecast, the disposition maps, GIS maps and transportation information. The system also allows users to customize and generate reports over the web. These reports can be exported to various formats, such as Adobe{sup R} PDF, Microsoft Excel{sup R}, and Microsoft Word{sup R} and downloaded to the user's computer. Future enhancements will include database/application migration to the next level. A new data import interface will be developed to integrate 2012-13 forecast waste streams. In addition, the application is updated on a continuous basis based on DOE feedback. (authors)

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

Identification and time evolution of thionyl chloride (SOCl 2 ) radiolysis products

In this study, innovative solutions are needed to reduce the amount of high-level waste generated by used nuclear fuel recycling strategies to support the widespread adoption of sustainable nuclear fission energy technologies. To this end, a new sulfur chloride-based process has been developed to recycle zirconium alloy-based materials, which make up a significant fraction of high-level radioactive waste. To support the continued development of this process, we present new data on the potential reaction pathways over time of the products arising from the gamma and electron beam radiolysis of neat thionyl chloride (SOCl 2 ). Interrogation of the gamma irradiated liquid by Raman spectroscopy provided more conclusive identification of the SOCl 2 degradation products, specifically sulfur dichloride (SCl 2 ), molecular chlorine (Cl 2 ), sulfur dioxide (SO 2 ), and sulfuryl chloride (SO 2 Cl 2 ). In comparison, the high dose rate (~10 7 Gy s -1 ) electron beam irradiations formed significantly more degradation products. For both cobalt-60 gamma and electron beam irradiations, the observed degradation products were found to evolve as a function of time post-irradiation via the same reaction pathways, with indication of a solvent regeneration mechanism. These findings are fortuitous for process development, as such a mechanism would be beneficial for process longevity and cost effectiveness.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

EnergySolution's Clive Disposal Facility Operational Research Model - 13475

EnergySolutions owns and operates a licensed, commercial low-level radioactive waste disposal facility located in Clive, Utah. The Clive site receives low-level radioactive waste from various locations within the United States via bulk truck, containerised truck, enclosed truck, bulk rail-cars, rail boxcars, and rail inter-modals. Waste packages are unloaded, characterized, processed, and disposed of at the Clive site. Examples of low-level radioactive waste arriving at Clive include, but are not limited to, contaminated soil/debris, spent nuclear power plant components, and medical waste. Generators of low-level radioactive waste typically include nuclear power plants, hospitals, national laboratories, and various United States government operated waste sites. Over the past few years, poor economic conditions have significantly reduced the number of shipments to Clive. With less revenue coming in from processing shipments, Clive needed to keep its expenses down if it was going to maintain past levels of profitability. The Operational Research group of EnergySolutions were asked to develop a simulation model to help identify any improvement opportunities that would increase overall operating efficiency and reduce costs at the Clive Facility. The Clive operations research model simulates the receipt, movement, and processing requirements of shipments arriving at the facility. The model includes shipment schedules, processing times of various waste types, labor requirements, shift schedules, and site equipment availability. The Clive operations research model has been developed using the WITNESS{sup TM} process simulation software, which is developed by the Lanner Group. The major goals of this project were to: - identify processing bottlenecks that could reduce the turnaround time from shipment arrival to disposal; - evaluate the use (or idle time) of labor and equipment; - project future operational requirements under different forecasted scenarios. By identifying processing bottlenecks and unused equipment and/or labor, improvements to operating efficiency could be determined and appropriate cost saving measures implemented. Model runs forecasting various scenarios helped illustrate potential impacts of certain conditions (e.g. 20% decrease in shipments arrived), variables (e.g. 20% decrease in labor), or other possible situations. (authors)

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

The Port Hope Areas Initiative's Port Granby Project: The Decontamination of Site Equipment from a Low Level Waste Facility - 20330

The Port Hope Area Initiative (PHAI) is the Government of Canada's response to the community recommended solutions for the remediation and safe, long-term management of historic low-level radioactive waste (LLRW) that was the result of uranium and radium processing operations in the Municipality of Port Hope, Ontario from the 1930's to the 1980's. The Eldorado refinery, on the shores of Lake Ontario, began refining radium-226 from pitchblende ore, later transitioning to the refining of uranium ore. Process residues were deposited at the Welcome Waste Management Facility in Port Hope until the mid-1950's switching to the Port Granby Waste Management Facility (PGWMF), which continued to accept waste until the late 1980's. The Port Granby Project involves the construction of a new, secure long-term waste management facility (LTWMF) and supporting infrastructure, and the excavation and relocation of approximately 1,200,000 m{sup 3} of historic low-level radioactive waste and marginally contaminated soil from the PGWMF to the newly constructed LTWMF. Since the project's inception in 2016, the bulk of the material has since been moved to the final long term storage cell. This has taken over 500,000 person-hours and a large amount of one site heavy equipment to accomplish including excavators, bulldozers, rock trucks, baker tanks and vacuum trucks. While many of the pieces of heavy equipment and water management equipment has further work to do on site, several of the sites larger items have already undergone the decontamination process prior to release from the site. This process has posed a serious challenge as the waste has a caustic nature to it and over the last few years it has had time to work its way into the metallic structures and framework of the equipment. Coupled with the tight work spaces and in some cases confined work areas there have been many issues that needed to be addressed to successfully reach a state where items can be free released to the unrestricted limits of the sites license. This ongoing process has led to a more refined decontamination program at the Port Granby Facility and has had a rather high success rate considering the length of time the equipment has been exposed to the waste materials on site. This paper discusses the varied challenges of decontamination faced at a low-level radioactive waste facility with a focus on the following aspects: - Challenges faced regarding different equipment types and the solutions implemented by Wood to address these challenges; - An examination of various decontamination methods employed throughout the project; and - Lessons learned over three years of remediation work and how they impact the decontamination process. (authors)

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

Waste Acceptance and Disposal at the Nevada National Security Site

The U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) relies on the Performance Assessment (PA) and the PA model of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada National Security Site (NNSS) to make disposal decisions on waste stream profiles submitted by waste generators. Disposal decisions are made following a formal process called the PA review process. This paper presents the PA review process, which is a part of the NNSA/NFO’s Waste Generator Documentation Approval Process. The review process includes three major steps and employs quantitative tools such as the PA-derived factors and probabilistic inventory and PA models developed in GoldSim. The PA-derived radionuclide action levels for waste characterization and reporting are included in the NNSS Waste Acceptance Criteria. Each year, NNSA/NFO reviews approximately 100 to 150 new or revised waste stream profiles. The PA review group ensures that, with the disposal of approved waste streams, the Area 5 RWMS will remain in compliance with the performance objectives of DOE O 435.1, “Radioactive Waste Management.” Examples of the PA review process are given that demonstrate how waste streams with special characteristics are evaluated and disposal decisions are made. The most rigorous profile evaluation is done with the PA model, run with the proposed profile inventory added and results compared to the PA results. Additional analysis and process-modeling are performed for those waste streams with unique features not addressed with the PA model. The results of these analyses and modeling are documented in special analysis reports, reviewed and approved by NNSA/NFO. The examples include (1) profiles with radionuclides not in the PA, (2) profiles including radionuclides with high radon-generation potential, and (3) profiles with high-heat generating radionuclides.

Yucel, Vefa↗
DOE OSTI2012

Radioactive Bench-scale Steam Reformer Demonstration of a Monolithic Steam Reformed Mineralized Waste Form for Hanford Waste Treatment Plant Secondary Waste - 12306

Hanford currently has 212,000 m{sup 3} (56 million gallons) of highly radioactive mixed waste stored in the Hanford tank farm. This waste will be processed to produce both high-level and low-level activity fractions, both of which are to be vitrified. Supplemental treatment options have been under evaluation for treating portions of the low-activity waste, as well as the liquid secondary waste from the low-activity waste vitrification process. One technology under consideration has been the THOR{sup R} fluidized bed steam reforming process offered by THOR Treatment Technologies, LLC (TTT). As a follow-on effort to TTT's 2008 pilot plant FBSR non-radioactive demonstration for treating low-activity waste and waste treatment plant secondary waste, TTT, in conjunction with Savannah River National Laboratory, has completed a bench scale evaluation of this same technology on a chemically adjusted radioactive surrogate of Hanford's waste treatment plant secondary waste stream. This test generated a granular product that was subsequently formed into monoliths, using a geo-polymer as the binding agent, that were subjected to compressibility testing, the Product Consistency Test and other leachability tests, and chemical composition analyses. This testing has demonstrated that the mineralized waste form, produced by co-processing waste with kaolin clay using the TTT process, is as durable as low-activity waste glass. Testing has shown the resulting monolith waste form is durable, leach resistant, and chemically stable, and has the added benefit of capturing and retaining the majority of Tc-99, I-129, and other target species at high levels. (authors)

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

Dismantling of the Spent Resins Hot Compaction Unit at Tihange NPP, Belgium - 19139

Design, construction and commissioning of the Spent Ion Exchange Resins Hot Compaction Unit at Tihange Nuclear Power Plant took place from 2005 till 2011. In this conditioning process, batches of radioactive resin beads were first dried to less than 4% moisture by a conical vacuum dryer. Resins were then poured in compactable 200-liter drums and super-compacted. Resulting pellets were emplaced in standard 400-liter drums, the packages qualified for final disposal in Belgium, and eventually grouted. The Belgian Waste Management Agency (ONDRAF/NIRAS) was consulted before initiating the project and issued a statement of no objection. The technology turned out to be used for the first time in the particular case of bead shaped resins (unlike powdered resins) and with a grouting requirement. Finally, commissioning tests ended in a safe and reproducible waste form, compliant with prevailing Waste Acceptance Criteria. Volume Reduction Factors of about 2 were recorded after drying and compaction. New tests were carried out to support the Qualification File of the process, which was submitted to ONDRAF/NIRAS. After performing many trials using inactive resins, the Operator dried two batches of radioactive resins and super-compacted four drums: this was a planned qualification step, aiming at the validation of process parameters. In the meantime, ONDRAF/NIRAS made parallel progress in the design of the future Near Surface Repository, where conditioned resins will be disposed of, and found some incompatibility between encapsulated dried resin beads and water intrusion hazards inside the repository. Based on the risk that dried resin beads will swell if they absorb water, causing mechanical stresses likely to compromise the integrity of the disposal site, ONDRAF/NIRAS informed Tihange NPP that the Hot Compaction Unit would not be qualified as spent resins conditioning process. Further to this irrevocable decision, the Operator undertook three actions: - The first one was the qualification and purchase of waste containers to provide additional temporary storage capacity on site for its operational spent resins; - The second one is the selection of a new spent resins conditioning process and the performance of the related qualification tests to demonstrate it meets the updated Waste Acceptance Criteria; - The last one is the dismantling of the Hot Compaction Unit to clear the space and enable the installation of new equipment. This latter task was carried out in 2018, involving decontamination and free release measurements of all the materials, a very limited fraction of it having to be further processed as radioactive waste. Nevertheless, Tihange NPP wishes to keep the super-compactor and its hydraulic unit for a possible resale or reuse for the treatment of mixed waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Metal-oxide cages open up strategy for processing nuclear waste

We report the ability to separate the radioactive element americium from spent nuclear fuel would lower the long-term hazards of nuclear waste. An inorganic molecular cage that selectively binds to americium opens up a separation strategy.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Blending as an In-Tank Waste Processing Strategy: Year 1 Progress Report

The Hanford Site stores approximately 56 million gallons of radioactive legacy defense waste in underground tanks. Direct feed of southeast quadrant tank waste to a high-level waste treatment facility is currently being explored to accelerate tank treatment activities. This study seeks to de-risk waste transfers by favorably altering the resuspension and critical velocities of wastes through strategic waste blending. Wastes rich in aluminum- and zirconium-bearing solids pose significant risks due to their high settled strength and density. A series of tests explores how solid phase particle size and suspending phase chemistry alter waste transport properties, quantified through the just-suspended mixing speed (NJS), for pure and blended aluminum- and zirconium-bearing waste solid simulants. The study utilizes a range of gibbsites (three sizes spanning 10 to 90 µm) and zirconia solids (three sizes spanning 10 to 100 µm). As expected, the NJS of pure solids largely follows predictions based on literature correlations; blends of gibbsite and zirconium-bearing solids of differing particle sizes exhibit NJS maxima attributed to dense packing of the mixed-size settled solids. The findings of this study will assist in developing strategies to optimize the transport of waste solids for the safe and efficient delivery of tank waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗