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MH-1A Sturgis Decommissioning and Dismantlement - 20281

The U.S. Army Corps of Engineers (USACE) with its prime contractor, Aptim Federal Services, LLC (APTIM), recycled 5,260 metric tons (11.6 million pounds) of material from Sturgis Barge (Sturgis) during the Decommissioning and Dismantlement of the MH-1A nuclear power reactor. The overall objective of the project was to reduce residual radioactivity associated with MH-1A to levels that permitted release of Sturgis for dismantlement and termination of the Army Reactor Office permit. By effectively applying waste hierarchy's three Rs - reduce, reuse and recycle - the Sturgis project not only minimized the amount of waste that required disposal at landfills, but also reduced the potential for long-term environmental liability emanating from these landfills. The project team completed the physical decommissioning efforts in June 2018 in Galveston, TX. In September 2018, radiological surveys were completed to demonstrate the vessel could be released for shipbreaking. Sturgis was towed from Galveston, TX to Brownsville, TX in late September 2018. Shipbreaking, dismantlement and recycling efforts began in early October 2018 and were completed on 15 March 2019. As part of the decommissioning effort in Galveston, the team shipped 69 shipments (860 metric tons) of low-level radioactive waste and radioactive components to the Waste Control Specialist (WCS) facility in Andrews TX for disposal. Certain radioactive components had to be transferred to the Department of Energy prior to placing the materials into the Federal Waste Facility located within WCS. However, most of the radioactive waste was characterized, profiled, approved and managed under the WCS permitted radioactive waste exemption process authorized and implemented by the Texas Commission on Environmental Quality (TCEQ) and the Radioactive Materials Division. This allows LLRW and LLMW to be shipped as regulated waste and then upon receipt at WCS through satisfying the relevant waste acceptance criteria the waste is exempted and placed into the WCS RCRA permitted cell. An additional 35 shipments (544 metric ton) of contaminated hazardous waste water were transported to U.S. Ecology in Robstown, TX for treatment/disposal. An additional 36 shipments (500 metric tons) of non-hazardous wastewater was sent to Republic Waste Services' facility in Fresno, TX. The disposal of these materials required close coordination with State of Texas regulators. During decommissioning, the project team recycled approximately 270 metric tons (600,000 pounds) of lead and steel. As part of the dismantlement in Brownsville, TX the team recycled approximately 5,000 metric tons (11 million pounds) of ferrous and non-ferrous material, limiting our disposal requirements to about 180 metric tons (400,000 pounds) of material (<4% from entire shipbreaking activity). Although the primary hazard being mitigated by this project was radiological, recycling was always a priority for the project. The team strived to achieve sustainability goals as we implemented this one of a kind project. Scrap metal recycling has a large positive impact on the environment and can also favorably impact project disposal costs. Steel is among the most recycled material in the world. Nearly 40% of the world's steel production is made from scrap. Recycling steel also requires 75% less energy than producing it from raw materials. By using recycled steel rather than virgin materials, 2.33 kg of carbon emissions are eliminated per kg of steel [1]. The project recycled more than 4,500 metric tons (10 million pounds) of steel, which eliminated about 10,400 metric tons (23 million pounds) of CO{sub 2}. By implementing a recycling initiative for the Sturgis project, the team was able to realize cost avoidance for disposal of scrap, cost savings from the metals recycled, plus the project provided benefits to the environment through our recycling efforts. Once the dismantlement was complete, the team prepared a detailed decommissioning closure report, which allowed for the termination of the Army Reactor Decommissioning Permit. While it not only reduced any potential long-term environmental liability, this project to decommission and dismantle a floating nuclear power plant is truly unprecedented - it is a prime example of the USACE mission which is: 'Engineering solutions for the Nation's toughest challenges'. This unique, one of a kind, historical power plant was never designed to be taken apart, and the available information about its construction was lacking in many details. The hazards that required mitigation dictated a painstaking and deliberate process in order to avoid any release to the environment and the community, and to protect the health and safety of the workers involved while keeping the waste hierarchy's three R's - reduce, reuse and recycle at the forefront. (authors)

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Waste Management Approach for a Potential Future U.S. Closed Fuel Cycle - 16233

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. 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. - 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 CaCO{sub 3} 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 titanate ceramic [e.g., 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 titanate ceramics. - 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 Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. - Electrochemical process salt HLW which is immobilized in 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 a representative flowsheet are reported. (authors)

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An Assessment of the Materials of Construction for the Transfer Lines and Unit Operations Equipment Associated with the Recycle Diversion Process

The Defense Waste Processing Facility (DWPF) processes and vitrifies radioactive waste that it receives from the Concentration, Storage, and Transfer Facility (CSTF) and Salt Waste Processing Facility (SWPF). As a result of the evaporation of water during both the melter feed preparation and the melter feed vitrification steps in DWPF, a recycle waste stream is generated, neutralized, and sent back to the CSTF. The recycle waste is a dilute water stream originating from the collection of condensate liquids containing some minor sludge and frit solids and other waste components primarily resulting from entrainment into the condensate during foam-over events. The recycle stream volume is significant and is expected to approach 3 million gallons per year once the SWPF reaches full operation. Diverting the bulk of the recycle waste stream from the CSTF is essential for the eventual closure of the waste tanks, and hence the completion of the SRS liquid waste mission.

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Pathways for Disposal of Commercially-Generated Tritiated Waste

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two of these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste transportation, processing and disposal vary based a number of factors. In many cases, wastes with very low radioactivity are priced primarily based on weight or volume. For higher activities, costs are based on both volume and activity, with the activity-based charges usually being much larger than volume-based charges. Other factors affecting cost include location, waste classification and form, other hazards in the waste, etc. Costs may be based on general guidelines used by an individual disposal or processing site, but final costs are established by specific contract with each generator. For this report, seven hypothetical waste streams intended to represent commercially-generated tritiated waste were defined in order to calculate comparative costs. Ballpark costs for disposition of these hypothetical waste streams were calculated. These costs ranged from thousands to millions of dollars. Due to the complexity of the cost-determining factors mentioned above, the costs calculated in this report should be understood to represent very rough cost estimates for the various hypothetical wastes. Actual costs could be higher or could be lower due to quantity discounts or other factors.

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Back-end of the fuel cycle - Indian scenario

Nuclear power has a key role in meeting the energy demands of India. This can be sustained by ensuring robust technology for the back end of the fuel cycle. Considering the modest indigenous resources of U and a huge Th reserve, India has adopted a three stage Nuclear Power Programme (NPP) based on 'closed fuel cycle' approach. This option on 'Recovery and Recycle' serves twin objectives of ensuring adequate supply of nuclear fuel and also reducing the long term radio-toxicity of the wastes. Reprocessing of the spent fuel by Purex process is currently employed. High Level Liquid Waste (HLW) generated during reprocessing is vitrified and undergoes interim storage. Back-end technologies are constantly modified to address waste volume minimization and radio-toxicity reduction. Long-term management of HLW in Indian context would involve partitioning of long lived minor actinides and recovery of valuable fission products specifically cesium. Recovery of minor actinides from HLW and its recycle is highly desirable for the sustained growth of India's NPPs. In this context, programme for developing and deploying partitioning technologies on industrial scale is pursued. The partitioned elements could be either transmuted in Fast Reactors (FRs)/Accelerated Driven Systems (ADS) as an integral part of sustainable Indian NPP. (authors)

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Contribution of the numerical simulations in the development of the French HLLW vitrification Processes - 15126

For many years, the CEA (Commissariat a l'energie Atomique et aux energies Alternatives) Marcoule France has developed various processes dedicated to radioactive waste confinement, especially vitrification processes for HLLW. For 10 years now, the numerical simulation has become an important tool for research and development projects held in the CEA-AREVA Joint Vitrification Laboratory (LCV). Induction heating, fluid mechanics and thermal simulations take part of all new R and D projects. The contributions of such simulations are, first, the enhancement of the working knowledge of existing process. Those data are very useful to define optimisation choices, for example upgrades made on the hot metallic melter used since the 90's at La Hague facility. Second, the simulations are, of course, also used at the conception stage of new processes as a tool allowing wide ranges parametric tests. This has been extensively used in the design of the cold crucible inductive melter (CCIM) commissioned in 2010 at La Hague plant. Finally, it is a powerful and relatively cheap tool for prospective studies for processes of the future. Whatever the purpose, the potential benefits are gains on the reliability, the output capacity and the life time. (authors)

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Fluor Idaho's Technical and Programmatic Approach That Resolved the Problems with the Integrated Waste Treatment Unit - 19290

The Integrated Waste Treatment Unit (IWTU) is a facility designed to treat approximately 900,000 gallons of radioactive liquid waste, referred to as Sodium Bearing Waste (SBW). IWTU is a Fluidized Bed Steam Reforming process that will be used to treat the SBW to produce a granular sodium carbonate and sodium aluminate product. The product is planned to be managed as remote handled transuranic waste. Construction of the IWTU was completed in 2011 and Critical Decision 4 was achieved in 2012 with the expectation that the treatment of the SBW would be completed in 2012. However, process instabilities and equipment problems prevented the safe and reliable operation of the facility. All of the nonradiological tests of the IWTU had to be stopped prematurely because each time the IWTU was operated, the process would become unstable within a few days of initiating the operation. The instabilities were associated with the primary fluidized bed reaction vessel, the Denitration Mineralization Reformer (DMR). The instabilities included difficulties maintaining fluidization conditions and scale formation within the DMR. Design inadequacies and equipment reliability problems in other parts of the facility also hampered the testing the IWTU. The U.S. Department of Energy (DOE) awarded the Idaho Clean-up contract to Fluor Idaho (Fluor) in 2016. To resolve the issues with the IWTU, Fluor assembled a dedicated team to develop the technical and programmatic strategy to resolve the issues with the IWTU and get it operational. Fluor's approach has successfully resolved the primary issues with the IWTU. As a result of modifications of equipment and operating strategies implemented by Fluor, the core part of the IWTU process now works reliably when processing simulated Sodium Bearing Waste over a wide range of operating conditions. This paper describes Fluor's approach and summarize key steps in the process to resolve the issues with the IWTU. Fluor has used strong fundamentals including a first principles approach and unique program management techniques to solve the technical problems with the IWTU. The successful approach has provided renewed confidence in the IWTU process enabling the troubled project to move forward. Important factors in the success of the project include using a dedicated project team separate from the IWTU line organization, a Technical Review Group, the Chemistry Team and a diverse set of supporting contractors and National Laboratories. Fluor typically used multiple means to validate solutions before testing them in the IWTU. Fluor's successful full-scale demonstration runs of the IWTU confirm the process will reliably produce a compliant, treated waste product over a wide range of operating conditions.

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Is Radioactive Tank Waste a Bingham Fluid?

Slurries and sludges across the United States Department of Energy (DOE) complex rank among the most rheologically interesting. Their composition is heterogeneous, spanning a very broad range of particle sizes, densities, and interparticle forces. All exhibit shear thinning, some have yield stresses, and many are thixotropic. Despite the variety, these complex fluids are often represented using the historic Bingham fluid model, which fits higher shear rate data to a simple straight line. The intercept provides a yield stress, which has been a key design parameter in construction of large-scale waste processing facilities. However, many radioactive wastes are simply not Bingham fluids, which extrapolate but poorly across low to intermediate shear rates that are characteristic of typical processing conditions. Indeed, processing shear rates as high as 200 1/s, which has been a typical minimum shear rate used in fitting, are seldom encountered in nuclear waste processing. Therefore, more realistic rheological models are necessary to accurately predict waste processing performance. PNNL recently re-evaluated the rheology of reconstituted Hanford REDOX sludge waste against a wide variety of rheological models including the Bingham, Cross, Cross with yield stress, Carreau, biviscous, Herschel-Bulkley (which includes a power law dependence), Casson, and Gay models. They found that all of the models provided a closer fit than the Bingham model and that the biviscous model and Cross with yield stress model were convincing. However, reconstituted Hanford REDOX sludge waste is but one type of DOE waste and a direct contrast and comparison of these three models against undiluted, unmixed tank waste (actual not simulant) has not been performed previously. Therefore, the purpose of this paper is to evaluate the rheology of actual tank waste with these more accurate rheological models. In this paper, we evaluate select rheological data for slurry samples from Hanford’s AZ-101, AZ-102, and SY-101 waste tanks. In each of these cases, we find that Cross’ model with yield stress and the biviscous model significantly outperform the Bingham fluid model. Furthermore, the AZ-101 data also shows that the shear stress peak that startup significantly exceeds the Bingham yield stress, which is commonly observed in the incipient moments of rheological measurements on simulants. In contrast, Cross’ model may empirically accommodate an initial spike in shear stress at modest shear rates. These are important observations because computational and analytical fluid dynamics simulations rely on rheological constitutive models for accurately and conservatively predict waste processing performance. These findings suggest the need for better rheological modeling of and validation against radioactive waste.

Rheology, Radioactive Waste↗

Effect of supplementary cementitious materials on the degradation of cement-based barriers in radioactive waste repository: A case study in Korea

This study focuses on investigating the chemical degradation characteristics of cementitious barriers used in low- and intermediate-level radioactive waste repository by reactive transport modeling. The impact of the blending with supplementary cementitious materials (SCMs) in the barriers on the chemical degradation was evaluated to find the optimum barrier design. A number of different barrier designs were examined by replacing ordinary Portland cement (OPC) by SCMs (i.e., fly ash, silica fume, and blast-furnace slag). The simulation results showed that silica fume blended barrier has better durability against chemical degradation by rainwater compared to fly ash or blast-furnace slag blended barriers. In addition, the chemical durability of silica fume-based barrier increased with increasing replacement level up to about 20 %. It seems that the amount of formed calcium silicate hydrate (CSH) in the initial cement-based barrier highly affects the overall chemical durability. The newly developed reactive transport model demonstrated its capability for understanding the barrier performance and investigating the optimal design of the barrier system.

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Summary of recent Raman Spectroscopy testing of SRS processes

This report describes several scoping projects conducted at SRNL using Raman spectroscopic methods for monitoring different aspects of nuclear waste and materials processing. One project examined the suitability of a Raman telescope for in situ measurement of solid residues in waste tanks. Characteristics evaluated for this equipment included radiation resistance, ease of use, and sensitivity. A second project monitored the nitrate content in liquid filtrate from the testing of a rotary microfilter using a fiber-based insertion probe. The third project made Raman measurements of various gases, including H 2 and NO x , in the headspace of a vessel while dissolving aluminum coupons in nitric acid. Measurements followed the evolution of these species in real time. Although the majority of these projects occurred in the laboratory environment, SRNL has substantial experience with implementing other optical techniques into nuclear materials processing environments. The work described in this report shows the potential of the Raman technology to provide real time measurements of species such as nitrate or hydroxide during sludge washing or evolved gases such as hydrogen or NO x during waste processing.

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Managing TRU waste processing and storage issues/challenges - 15487

Operational events at the Waste Isolation Pilot Plant (WIPP) facility resulted in suspension of disposal operations and waste shipments to WIPP creating significant challenges for the Department of Energy (DOE) complex. The Oak Ridge Transuranic (TRU) Waste Program which was poised to re-start shipping after a 2.5 year suspension of the Carlsbad Field Office (CBFO) certification activities which resulted in on-site accumulation of the highest activity contact handled (CH) TRU waste containers. As a result, Transuranic Waste Processing Center (TWPC) is nearing its physical storage capacity for CH waste and has already reached storage capacity limits for remote handled (RH) TRU waste. The TWPC storage capacity for RH-TRU waste is very limited and dependent on real time shipping support to continue RH waste processing activities in the facilities Hot Cell. Additionally, changes in dispersion modeling (i.e., dry deposition factor changes) used in the TWPC Documented Safety Analysis (DSA) present Material at Risk (MAR) management challenges to effectively manage off-site dose consequences. These facts necessitated the identification of alternative waste storage options for both CH and RH waste to continue waste processing operations at TWPC. (authors)

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Material Properties and Temperature Effects Quantified for Slurry Wear - 16141

Literature indicates that the hardness of solid particles relative to the target material hardness and temperature influence erosion. Investigations of temperature effects on erosion also show that erosion resistance is a function of target material's homologous temperature to melting point temperature. While relative hardness effects on erosion and temperature effects on erosion have been investigated independently, their combined effect has not been evaluated. This paper summarizes a review of relevant literature on relative hardness and temperature effects on erosion and via probabilistic calculations, quantifies the combined effect of relative hardness and temperature on liquid-particle induced wear, or slurry wear. These insights can assist material selection and plant operations at waste processing facilities. (authors)

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Microwave processing of cement and concrete materials – towards an industrial reality?

Each year a substantial body of literature is published on the use of microwave to process cement and concrete materials. Yet to date, very few if any have lead the realisation of a commercial scale industrial system and is the context under which this review has been undertaken. The state-of the–art is evaluated for opportunities, and the key barriers to the development of new microwave-based processing techniques to enhance production, processing and recycling of cement and concrete materials. Applications reviewed include pyro-processing of cement clinker; accelerated curing, non-destructive testing and evaluation (NDT&E), and end-of-life processing including radionuclide decontamination.

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DWPF Safely Dispositioning Liquid Waste

The only operating radioactive waste glassification plant in the nation, the Defense Waste Processing Facility (DWPF) converts the liquid radioactive waste currently stored at the Savannah River Site (SRS) into a solid glass form suitable for long-term storage and disposal. Scientists have long considered this glassification process, called “vitrification,” as the preferred option for treating liquid radioactive waste.

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Hanford Low-Activity Waste Processing: Demonstration of the Off-Gas Recycle Flowsheet - 13443

Vitrification of Hanford Low-Activity Waste (LAW) is nominally the thermal conversion and incorporation of sodium salts and radionuclides into borosilicate glass. One key radionuclide present in LAW is technetium-99. Technetium-99 is a low energy, long-lived beta emitting radionuclide present in the waste feed in concentrations on the order of 1-10 ppm. The long half-life combined with a high solubility in groundwater results in technetium-99 having considerable impact on performance modeling (as potential release to the environment) of both the waste glass and associated secondary waste products. The current Hanford Tank Waste Treatment and Immobilization Plant (WTP) process flowsheet calls for the recycle of vitrification process off-gas condensates to maximize the portion of technetium ultimately immobilized in the waste glass. This is required as technetium acts as a semi-volatile specie, i.e. considerable loss of the radionuclide to the process off-gas stream can occur during the vitrification process. To test the process flowsheet assumptions, a prototypic off-gas system with recycle capability was added to a laboratory melter (on the order of 1/200 scale) and testing performed. Key test goals included determination of the process mass balance for technetium, a non-radioactive surrogate (rhenium), and other soluble species (sulfate, halides, etc.) which are concentrated by recycling off-gas condensates. The studies performed are the initial demonstrations of process recycle for this type of liquid-fed melter system. This paper describes the process recycle system, the waste feeds processed, and experimental results. Comparisons between data gathered using process recycle and previous single pass melter testing as well as mathematical modeling simulations are also provided. (authors)

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Modeling of Coupled Thermo-Hydro-Mechanical-Chemical Processes for High-Level Radioactive Waste Repositories - 17361

To ensure that multi-barrier systems in high-level radioactive waste (HLW) repositories can efficiently perform their functions and isolate the waste from the biosphere, coupled thermo-hydro-mechanical-chemical (THMC) processes used for the performance assessment of HLW disposal facilities should be modeled. To evaluate coupled THMC processes, the DECOVALEX project, a unique international research collaboration project initiated in 1992, developed mathematical models and field experiments involving THMC processes in engineered barrier systems and host rocks for HLW disposal. The DECOVALEX project and the current status of coupled THMC model development can provide further insights into the performance assessment of HLW disposal facilities. This study provides information regarding the DECOVALEX project and reviews the development of coupled THMC models. The study results furnish knowledge of coupled THMC processes and techniques for the performance assessment and safety analysis of the final HLW disposal. (authors)

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Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

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Dewatering Innovations - 16381

The Liquid and Gaseous Waste Operations (LGWO) project, located at the Oak Ridge National Laboratory, handles wastewater from over 600 generators in 200 facilities. LGWO processes these waters, removing radioactive and chemical contaminants prior to discharge to the environment. Through these processes, LGWO generates various sludges that require dewatering prior to transportation. The original process was very inefficient. It involved placing PVC tubing in the bottom of a 90 ft{sup 3} box and sluicing the sludge into the box. Suction was then applied to the PVC tubes to remove water that would settle out of the material. The old process took several evolutions to drain/decant the liquids, which caused unnecessary worker radiation exposure. The moisture content was Near High Moisture content, which presented the possibility that liquids could be released during transport. Due to this fact, disposal costs were very high. Solutions for the 'Old' disposal process were to use liquid-rated Super Sacks, add additional Absorbent and then have the containers evaluated for liquid at TSDRF, prior to final disposal. This paper will describe the innovative packaging that was developed for the processing of these sludges, the ease with which the process was implemented, the versatility of the system and the cost savings associated with the new process. This system can be used in many other water treatment applications and erosion control. (authors)

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