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Characterization of oxide films on aluminum materials following reactor exposure and wet storage in the SRS L-Basin

This report describes the characterization of oxide films on one cropped piece of 6061, one 6063, and one cropped piece that could have been either 6061 or 6063 aluminum-based alloy, stored wet for up to approximately 40 years in L-Basin on the Savannah River Site (SRS). One 6061 cropping was from a Missouri University Research Reactor (MURR) fuel element. The 6063 piece was from a cropping from a Universal Sleeve Housing (USH). The Mark-16B fuel assembly could have been either 6061 or 6063 based on design drawings. Both the Mark-16B and USH were used in SRS Production Reactors. Characterization of the as-received oxides includes the oxide morphology, thickness, structure, and chemical composition. The aim is to characterize the oxide layer on aluminum cladding and fuel components that have reactor service and long-term basin storage service.

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Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

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Deep Geologic Nuclear Waste Disposal - No New Taxes - 12469

To some, the perceived inability of the United States to dispose of high-level nuclear waste justifies a moratorium on expansion of nuclear power in this country. Instead, it is more an example of how science yields to social pressure, even on a subject as technical as nuclear waste. Most of the problems, however, stem from confusion on the part of the public and their elected officials, not from a lack of scientific knowledge. We know where to put nuclear waste, how to put it there, how much it will cost, and how well it will work. And it's all about the geology. The President's Blue Ribbon Commission on America's Nuclear Future has drafted a number of recommendations addressing nuclear energy and waste issues (BRC 2011) and three recommendations, in particular, have set the stage for a new strategy to dispose of high-level nuclear waste and to manage spent nuclear fuel in the United States: 1) interim storage for spent nuclear fuel, 2) resumption of the site selection process for a second repository, and 3) a quasi-government entity to execute the program and take control of the Nuclear Waste Fund in order to do so. The first two recommendations allow removal and storage of spent fuel from reactor sites to be used in the future, and allows permanent disposal of actual waste, while the third controls cost and administration. The Nuclear Waste Policy Act of 1982 (NPWA 1982) provides the second repository different waste criteria, retrievability, and schedule, so massive salt returns as the candidate formation of choice. The cost (in 2007 dollars) of disposing of 83,000 metric tons of heavy metal (MTHM) high-level waste (HLW) is about $ 83 billion (b) in volcanic tuff, $ 29 b in massive salt, and $ 77 b in crystalline rock. Only in salt is the annual revenue stream from the Nuclear Waste Fund more than sufficient to accomplish this program without additional taxes or rate hikes. The cost is determined primarily by the suitability of the geologic formation, i.e., how well it performs on its own for millions of years with little engineering assistance from humans. It is critical that the states most affected by this issue (WA, SC, ID, TN, NM and perhaps others) develop an independent multi-state agreement in order for a successful program to move forward. Federal approval would follow. Unknown to most, the United States has a successful operating deep permanent geologic nuclear repository for high and low activity waste, called the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. Its success results from several factors, including an optimal geologic and physio-graphic setting, a strong scientific basis, early regional community support, frequent interactions among stakeholders at all stages of the process, long-term commitment from the upper management of the U.S. Department of Energy (DOE) over several administrations, strong New Mexico State involvement and oversight, and constant environmental monitoring from before nuclear waste was first emplaced in the WIPP underground (in 1999) to the present. WIPP is located in the massive bedded salts of the Salado Formation, whose geological, physical, chemical, redox, thermal, and creep-closure properties make it an ideal formation for long-term disposal, long-term in this case being greater than 200 million years. These properties also mean minimal engineering requirements as the rock does most of the work of isolating the waste. WIPP has been operating for twelve years, and as of this writing, has disposed of over 80,000 m{sup 3} of nuclear weapons waste, called transuranic or TRU waste (>100 nCurie/g but <23 Curie/1000 cm{sup 3}) including some high activity waste from reprocessing of spent fuel from old weapons reactors. All nuclear waste of any type from any source can be disposed in this formation better, safer and cheaper than in any other geologic formation. At the same time, it is critical that we complete the Yucca Mountain license application review so as not to undermine the credibility of the Nuclear Regulatory Commission and the scientific community. (authors)

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Accelerating the Decommissioning of the Sellafield Pile Fuel Storage Pond Using Innovative Remote Tooling Developed with the Supply Chain - 16437

The Pile Fuel Storage Pond (PFSP) at Sellafield and adjoining de-canning building provided the storage, cooling and de-canning facility for used fuel and isotopes from the two Windscale reactors and later some Magnox fuel from the Calder Hall reactors. Operations ceased in the 1970's and the facilities are now undergoing a programme of accelerated decommissioning. The pond is a sub-divided outdoor storage pond containing skips of irradiated fuel and miscellaneous waste items and the building contains a series of sub-ponds, known as bays, connected underwater to the main pond. The radiation environment and access limitations necessitate the use of remote technology for the decommissioning programme. This paper describes the philosophy and approach and the technology and equipment adopted for the removal and export of key plant equipment and waste items from within the PFSP. In order to progress the project cost effectively and speedily, the Sellafield project team looked for partners in the supply chain with the required experience, expertise and facilities to supply and test 'commercial off-the-shelf' (COTS) or modified proprietary equipment. This COTS approach minimised the need for multiple, expensive and lengthy design and manufacturing tasks associated with bespoke equipment. Sellafield partnered with specialist remote handling contractor, James Fisher Nuclear, and the respective project teams worked closely together at every stage of the project from qualifying requirements, identifying potential solutions, through to development, testing, training and final deployment. Optioneering of potential solutions involved early 'proof of principle' trials in JFN rig hall facilities, utilising existing equipment and plant mock-ups, along with other equipment manufactured in-house. Together the project teams selected the best solution to take forward, adopting a practical and trials based approach, which allowed rapid and cost effective modification of the solution throughout the project with minimal impact on the programme. When the final solution was validated, plant operators then gained familiarity and training on the equipment so they could execute the work efficiently on plant and so minimise their dose uptake. The removal of a Large De-canner is described in detail. This was a significant obstruction to the emptying, modification and re-use of the bay and a key milestone in the clean-up of the PFSP. The de-canners are large and heavy pieces of equipment which weigh 6.5 t each and approximately 6.5 m long. Their removal was further complicated by limited equipment and man access, radiological contamination from de-canning fuel, challenging material properties, unknown equipment condition and complex lifting requirements. Sellafield Ltd and JFN project teams worked closely together to rapidly develop, trial and deploy a succession of adapted 'commercial off-the-shelf' (COTS) tools whilst addressing emerging challenges and completing the overall task on time and budget. The Sellafield and JFN project and design teams provided an underwater: wheel stop removal tool; towing trolley; grinder on a carbon fibre pole, chute cutting 600 mm circular saw and hydraulic 6 t winch. The suite of tools will be reused on the remaining de-canners in the PFSP and First Generation Magnox Fuel Storage Pond (FGMSP) giving cost and programme savings. Each of the tools can also be used, and some have already been, for other tasks in the bay retrievals such as the size reduction of support frames, large steel sections and chemical de-canner preparatory works. The 'learning from experience' acquired during the project was shared by the Sellafield project with other decommissioning projects across the site. (authors)

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Waste management best practices for decommissioning of nuclear power plants - 15274

Decommissioning of a nuclear power plant involves the safe disposition of a large quantity of radioactive, hazardous and conventional waste. A number of nuclear power plant decommissioning projects have been completed or are nearing completion in the US, Germany and Spain. These projects have used various waste disposal and interim storage options, and have developed successful methods for handling the large quantities of waste created by the decommissioning. Additionally, decommissioning projects in several Asian and European countries are in progress or are in the planning stages such that strategies for the handling of decommissioning waste are being developed and/or implemented to address the regulatory requirements and disposal options available in these countries. These experiences provide important lessons learned and successful strategies for the handling of waste from power plant decommissioning projects. As the handling and disposition of decommissioning waste have a substantial effect on power plant decommissioning cost, schedule and complexity, these lessons learned are valuable for planning future decommissioning projects. (authors)

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SRNL Training Courses for Packaging, Transportation, and Storage of Radioactive Material - 17597

The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation has sponsored a suite of training courses that have been conducted since the early 1990s. Of these training courses, the Savannah River National Laboratory (SRNL) is currently leading the effort for two courses: (1) Management of SARP Preparation, and (2) Nuclear and Other Radioactive Materials Transport Security and Leak Testing. This paper provides an overview of both courses. (authors)

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Glass Formulation Testing to Increase Sulfate Incorporation

About 50 million gallons of high-level mixed waste is currently in storage in underground tanks at The United States Department of Energy's (DOE's) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE's Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal.

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The High Capacity Type B Waste Transport Cask to Serve Dismantling Project - 18344

The United States have already announced around 20 reactor shutdown and many more in the next 10 year. Financial constraints on plant operators and the low cost of fossil energy contribute to accelerate the shutdown of additional reactors. The industry has been gaining experience with reactor dismantling and is continuously trying to improve the cost efficiency of the upcoming projects. To do so, larger and more capacitive transport casks are needed to transport increasing volume of Low Level Waste (LLW) at a more attractive cost. Waste disposal is also a major cost element in a dismantling project. Consequently, the capacity to segment, package and transport the waste in the most optimized configuration will result in significant cost savings. While for LLW, the constraints for package requirements are limited, activated waste resulting mainly from reactor vessel and vessel internals segmentation require additional shielding to meet disposal and transportation regulations. The key element to succeed is the availability on the market of a licensed type B transport package with a larger payload that allows at the same time an optimization of the waste segmentation. AREVA TN has designed and licensed versatile high capacity type B casks for the transport of spent fuel canisters as part of the NUHOMS{sup R} system that can also be used to transport radioactive waste. While it will take a few more years to transport spent fuel to a consolidated interim storage facility and even more to a repository, the type B casks can be used to transport activated LLW from dismantled reactors to disposal sites. A first of a kind project will be performed with AREVA's partner NorthStar at Vermont Yankee Nuclear Power Plant. AREVA's team will perform the reactor segmentation and package the material into an AREVA's large capacity type B cask. This unmatched cask capacity is expected to significantly decrease the LLW disposal costs, as well as the onsite operations costs, while optimizing the schedule. The package will be transported by rail to waste disposal facilities. This paper will present the main characteristics of the package, provide an update of the overall schedule, and explain the innovations that contribute to simpler and more cost effective transport and operation. Finally, it will emphasize the benefits of using larger type B casks for immediate and future transportation needs. (authors)

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Decovalex-2019 (Executive Summary)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. The overall aim of DECOVALEX-2019 was to increase the understanding of various thermo-hydro-mechanical-chemical processes of importance for radionuclide release and transport from a repository to the biosphere and how they can be described and modelled using mathematical models. The scientific and technical objectives are: to increase the basic understanding of T-H-M-C coupled processes in fractured rocks (crystalline, sedimentary, argillaceous) and buffer materials; to investigate the predictive capabilities of different codes to field experiments and to perform verification of codes; to exchange experimental data, and improve the understanding of the constitutive behavior of crystalline and argillaceous rock masses and buffer materials; and to perform THMC calculations in a performance/safety assessment context.

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DECOVALEX-2019 (Task E Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

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Design and Materials of Reference Electrodes for Radioactive Waste Tank Service – A Literature Review

The Hanford site stores approximately 55 million gallons of radioactive and chemically hazardous wastes from the production of weapons materials. The wastes are stored in 177 underground, carbon steel storage tanks, 149 of these are single shell tanks (SSTs) and 28 of these are double shell tanks (DSTs). The DSTs provide critical retrieval and interim storage before the waste is vitrified in the Waste Treatment and Isolation Plant (WTP). The DSTs have been in service for 38 to 56 years and current plans indicate that WTP operations will be completed in 2075. Thus, the tanks will need to remain in service far beyond the initial 40-year life expectancy. For life extension of the tanks, effective corrosion control practices must remain in force. This effort includes direct measurements of the extent of corrosion (e.g., ultrasonic measurements and corrosion coupons) and electrochemical processes (e.g., linear polarization measurements).

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Characterization Report for Strontium Titanate in SWSA 7 And Adjacent Parcels in Support of the National Priorities List Site Boundary Definition Program, Oak Ridge, Tennessee

The U.S. Department of Energy (DOE) Oak Ridge Office requested support from the Oak Ridge Institute for Science and Education (ORISE) contract to delineate the extent of strontium titanate (SrTiO3) contamination in and around Solid Waste Storage Area (SWSA) 7 as part of the Oak Ridge National Priorities List Site boundary definition program. The study area is presented in Fig. 1.1 relative to the Oak Ridge Reservation (ORR). The investigation was executed according to Sampling and Analysis Plan/Quality Assurance Project Plan (SAP/QAPP) (DOE 2011) to supplement previous investigations noted below and to determine what areas, if any, have been adversely impacted by site operations.

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Enhanced safety of geologic CO 2 storage with nanoparticles

In this paper, some methods have been developed to detect leakage of CO 2 from its desired storage domain, but that is not sufficient to prevent and mitigate a leak. Two techniques have been proposed to prevent the migration of buoyant CO 2 from the storage domain by expediting mixing of CO 2 with the brine and mitigate risk of its leakage risk. These two methods are injection of CO 2 pre-mixed with brine, and injection of CO 2 with nanoparticles (NPs). The former has been studied to some extent, however, understanding of the latter is very limited. Unlike the application of NPs in hydrocarbon recovery, its use to enhance safety of CO 2 storage is a fairly unexplored topic that can have important benefits for the safety of the storage process. Also, the use of NPs for subsurface application in general is compromised for its cost. We investigate how NPs produced from low-level nuclear waste can be added with injected CO 2 to enhance the mixing of CO 2 with brine, which can mitigate leakage risk of CO 2 .

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Managing Risks via Small/Pilot Scale Testing for INL's Integrated Waste Treatment Unit (IWTU) - 19268

In 2005, the THOR steam reforming fluidized-bed reactor (FBR) technology was selected by the U.S. Department of Energy (DOE) as the technology of choice for treatment of sodium- bearing waste (SBW) at the INL. The basis of the process was to use fluidized bed technology to convert an acidic radioactive nitrate-rich liquid waste into a solid carbonate product suitable for long-term disposal and storage. During initial commissioning, as well as subsequent start-ups, the plant faced many operational challenges, such as stagnant/de-fluidized bed, and wall-scale formation. Such operational issues prompted the original contractor to seek expertise in pilot-scale testing, thereby reaching out to Hazen Research Inc. (HRI, Golden, Co). Several tests were conducted, but the results did not seem to address the underlying issues observed in the full-scale plant, as evidenced by continued operational issues. A systematic approach was then taken by the Fluor-Idaho team (new contractor as of June 2016) with respect to solving these technical issues. First, two different bench-scale, 2-inch fluidized bed reactors (stainless-steal and quartz), were utilized to test several theories to better understand the underlying chemical reactions that may be taking place under expected normal operating conditions. Then, the team designed and built an 18-inch FBR unit, mimicking the full-scale main vessel at IWTU - Denitration Mineralization Reformer (DMR). This was used to test various aspects of operational challenges observed at IWTU, namely particle size control, wall-scale growth-rate control, long-term operational stability, process optimization, and process off- gas filtration challenges. An example of information obtained from bench-scale testing, which was then tested during pilot-scale testing, was the phenomena that an addition of 5 vol % of CO{sub 2} gas into the fluidizing gas system nearly eliminated the formation of agglomerates that had typically led to bed de-fluidization. This change proved to be correct in both the pilot and full-scale testing and will be implemented as part of a long-term operational strategy of IWTU. Several other findings that have provided answers with respect to IWTU operational issues will be discussed in this report. (authors)

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Radioactive Water Treatment at a United States Environmental Protection Agency Superfund Site - 12322

A water treatment system at a United States Environmental Protection Agency (USEPA) Superfund site impacted by radiological contaminants is used to treat water entering the site. The United States Army Corps of Engineers (USACE) is actively managing the remedial action for the USEPA using contracts to support the multiple activities on site. The site is where former gas mantle production facilities operated around the turn of the century. The manufacturing facilities used thorium ores to develop the mantles and disposed of off-specification mantles and ore residuals in the surrounding areas. During Site remedial actions, both groundwater and surface water comes into contact with contaminated soils and must be collected and treated at an on-site treatment facility. The radionuclides thorium and radium with associated progeny are the main concern for treatment. Suspended solids, volatile organic compounds, and select metals are also monitored during water treatment. The water treatment process begins were water is pumped to a collection tank where debris and grit settle out. Stored water is pumped to a coagulant tank containing poly-aluminum chloride to collect dissolved solids. The water passes into a reaction tube where aspirated air is added or reagent added to remove Volatile Organic Compounds (VOC'S) by mass transfer and convert dissolved iron to a solid. The water enters the flocculent polymer tank to drop solids out. The flocculated water overflows to a fluidized bed contact chamber to increase precipitation. Flocculation is where colloids of material drop out of suspension and settle. The settled solids are periodically removed and disposed of as radioactive waste. The water is passed through filters and an ion exchange process to extract the radionuclides. Several million liters of water are processed each year from two water treatment plants servicing different areas of the remediation site. Ion exchange resin and filter material are periodically replaced and disposed of as radioactive waste. A total of 0.85 m{sup 3} of waste sludge per year requires disposal on average, in addition to another 6.6 m{sup 3} of waste cartridge filters. All water discharges are regulated by a state of New Jersey Pollutant Discharge Elimination System Permit implemented by the Federal Water Pollution Control Act (Clean Water Act). Laboratory analyses are required to satisfy requirements of the state NPDES permit. Specific monitoring parameters and discharge rates will be provided. Use of the water treatment systems drastically reduces the amount of contaminated water requiring solidification and water disposal to near zero. Millions of liters of potentially contaminated water from excavation activities is treated and released within permit limits. A small volume of solid radioactive waste (21 cubic meters) is generated annually from water treatment process operations. Management of ground and surface water is effectively controlled in remediation areas by the use of sumps, erosion control measures and pumping of water to storage vessels. Continued excavations can be made as water impacting the site is effectively controlled. (authors)

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History of Ultrasonic Inspection of High-Level Waste Tanks at SRS - 20220

The Liquid Waste Disposition Projects (LWDP) Structural Authority and Inspection Engineering Support Group has teamed with the Materials Science and Technology Non-destructive Evaluation Group of Savannah River National Laboratory (SRNL) to implement a comprehensive In-Service Inspection (ISI) program for the High-Level Waste (HLW) tanks in the sites tank farms. The purpose of this program is to demonstrate the structural and leak integrity of the tanks to maintain the function of waste containment throughout the desired service life. Four types of waste tanks were constructed at Savannah River Site (SRS) from 1951 thru 1981 to store radioactive nuclear waste from the separations process. The Type I, II and III are carbon steel cylindrical tanks with varying degrees of access to the side walls and very limited access to the tank bottom. The tanks contain 750,000 to 1.3 million gallons of waste and are buried or backfilled with soil. The tanks are 75'-85' in diameter and 24'-33' from tank bottom to the tank roof. Inspection of the HLW tanks is made difficult by radiation, contamination, and small access openings. Through the years these challenges have been overcome by progressive improvements to programs and equipment resulting in the sophisticated wall crawler Ultrasonic Testing (UT) technology of today. A formal inspection program was developed and put in service in 1972. UT thickness measurements were first made in 1967 and 1969 of selected waste tanks using an analog-type instrument to measure the thickness of the carbon steel walls. Routine inspections for thinning were performed from 1972-1985 and then discontinued because no thinning had been observed and this technique could not detect pitting or cracking. UT inspections resumed in 1994 using a sophisticated robotic wall crawler with UT technology capable of detecting thinning, cracking and pitting. The current program focuses on the newer Type III tanks that are used primarily for storage. The program was upgraded in 2003 to include selected regions of the tank wall, secondary wall and annulus floor on a 7 to 10-year frequency. The facility is currently in its third cycle of tank inspections. Thus far, there have been no reportable indications of service induced degradation. This result provides assurance that the chemistry control program has been effective at mitigating corrosion and that the tanks remain structurally sound even after more than 40 years of service. This paper will present the UT technology first used at SRS and the evolution to the current systems and their capabilities. Additionally, plans for equipment upgrades that will improve the extent of inspection will be discussed. (authors)

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Changes in the pore network structure of Hanford sediment after reaction with caustic tank wastes

At the former nuclear weapon production site in Hanford, WA, caustic radioactive tank waste leaks into subsurface sediments and causes dissolution of quartz and aluminosilicate minerals, and precipitation of sodalite and cancrinite. This work examines changes in pore structure due to these reactions in a previously-conducted column experiment. The column was sectioned and 2D images of the pore space were generated using backscattered electron microscopy and energy dispersive X-ray spectroscopy. A pre-precipitation scenario was created by digitally removing mineral matter identified as secondary precipitates. Porosity, determined by segmenting the images to distinguish pore space from mineral matter, was up to 0.11 less after reaction. Erosion-dilation analysis was used to compute pore and throat size distributions. Images with precipitation had more small and fewer large pores. Precipitation decreased throat sizes and the abundance of large throats. These findings agree with previous findings based on 3D X-ray CMT imaging, observing decreased porosity, clogging of small throats, and little change in large throats. However, 2D imaging found an increase in small pores, mainly in intragranular regions or below the resolution of the 3D images. Also, an increase in large pores observed via 3D imaging was not observed in the 2D analysis. Changes in flow conducting throats that are the key permeability-controlling features were observed in both methods.

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