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At least 253 records · Page 14

Test Bed: Treatability of Hanford Tank Waste - 18501

In an effort to advance the Low Activity Waste Pretreatment System technology readiness level at the Hanford Site, the effectiveness of a critical technology element, sRF ion-exchange resin, was demonstrated in the selective removal of Cs-137 from 11.5 L of composited Hanford supernate tank waste. Prior to ion-exchange, solids filtration was performed on the tank waste composite. Between the time of filtrate collection and filtrate processing on the ion-exchange columns, additional solids had formed. These secondary solids resulted from the combination of supernates high in fluoride content with supernates high in phosphate content and the relatively short blend time. Samples were collected before, during and after the ion-exchange process for characterization of a full suite of constituents (inorganic, organic, and radionuclides) to aid in the classification of the cesium-removed waste for shipping, receiving, treatment, and disposal determinations. Following chemical analysis, the remaining 10.8 L (2.9 gal) of cesium-depleted waste was shipped to an off-site vendor for grout-encapsulation and eventual disposal at the Waste Control Specialists Federal Waste Disposal Facility in Andrews, Texas as low-level waste. (authors)

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Challenging Engineering Demographics and the READY Program - 18528

The mission of the Liquid Waste Organization at the Savannah River Site involves the safe storage of nuclear waste, closure of waste tanks, and immobilization of the waste into grout (low-level waste) or glass (high-level waste). With much of the mission left to complete and a significant percentage of the engineering organization eligible for retirement within the next 5 years, focus has been on recruiting and retaining new hires and early career engineers. Over the past two years, enhancements have been made to current programs and new programs have been created with the goal of improving the retention rate for engineering. These programs include the Reaching Engineers at the Development Years, the Early Career Professional Technology Exchange, the Shift Technical Engineer Pipeline, and Technical Staff Training. (authors)

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Environmental Management Waste Management Facility Operations Team Commitment to Safety - 18546

The Environmental Management Waste Management Facility (EMWMF) is an above-grade waste disposal facility that is owned by DOE and operated by URS | CH2M Oak Ridge LLC (UCOR), a contractor to DOE's Oak Ridge Office of Environmental Management. The EMWMF landfill is located on the Oak Ridge Reservation (ORR) and has a 2.18 M cubic yard capacity. EMWMF is a key component of the cost-effective ORR cleanup program and accepts remediation waste from across the ORR. Multiple low-level and low-level mixed CERCLA waste streams are disposed at EMWMF, mostly as direct-dumped demolition and remediation debris. EMWMF receives and disposes of a myriad of contaminated waste streams and waste types, including massive structural steel from some of the largest buildings ever constructed, contaminated soils from legacy disposal areas, containers that must be grouted under negative pressure, and special waste that must be tightly controlled. Support activities include managing several million gallons of landfill wastewater per year and driving around 20,000 miles per year to transport leachate to a treatment facility. Even with the wide range of waste accepted and the diversity of support activities, EMWMF Operations has maintained an impressive safety record of no lost workday cases in nearly 16 years of continuous operation and was recently awarded the National Safety Council's Superior Safety Performance Award. This safety record is a testament to the commitment to safety by both workers and management. Several factors are key to the EMWMF Operations safety record. UCOR senior management established safety as a priority and allocates proper funding and resources. EMWMF Operations management embraces safety as an integral part of the work and also 'walks the talk'-emphasizing safety over productivity and showing personal commitment to safety. A stable workforce promotes a culture where safety is valued and helps train new personnel in this culture. The UCOR integrated safety management approach to work control is fully embraced, where worker feedback is highly valued and used to better identify and control job hazards. Finally, job supervisors actively encourage input. Worker concerns, suggestions, and observations are encouraged and positively addressed. As a result, the entire Operations Team is committed to identifying and mitigating hazards. This team commitment to safety results in a culture where safety is not only valued, but is expected. Safety concerns are identified and addressed proactively, and communication is encouraged. This strong safety culture and outstanding safety record have also proven to be good business, with the facility available to receive waste 99.6% of available work hours and outstanding turnaround times on waste shipments. The EMWMF Operations team commitment to safety can be used in other environments to develop an engaged, safety-conscious workforce. (authors)

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Radioactive Test Platform Demonstrating the Direct Feed Low-Activity Waste Process at Hanford - 18628

The Hanford Site is implementing a phased process for treating liquid legacy radioactive waste. The process is called the Direct Feed Low-Activity Waste process (DFLAW). The DFLAW includes a filtration and ion-exchange unit in the Low-Activity Waste Pretreatment System (LAWPS) that removes solids and cesium. This pretreat feed is then sent to a vitrification facility where glass-forming solids are added and the waste is turned into glass in a melter. The melter off-gas liquid condensate is evaporated in the Effluent Management Facility (EMF), with the concentrate recycled back into the feed. A laboratory-scale test platform has been built to test various aspects of the DFLAW process using samples of actual radioactive waste. The current platform represents four of the major unit operations: filtration, ion-exchange, vitrification, and off-gas condensation. A lab scale EMF is being designed to supplement the current platform. The platform is expandable to add any or all unit operations. The radioactive test platform has three missions. The first is simply to demonstrate the DFLAW process with real waste. A second objective is to address flowsheet gaps and opportunities, some of which require testing with real waste. Some gaps and opportunities include addressing the fate of mercury in the flowsheet, evaluating salt and aluminum precipitation, and proving new glass formulations with real waste. A third objective of the radioactive test platform is to demonstrate or evaluate future operations and alternative flowsheets. Some of the future operations evaluated will be the life expectancy of the resin in real waste, resin sampling, evaluation of how waste affects melter throughput, and grouting various primary and secondary waste streams. The radioactive test platform is therefore a crucial tool for verifying and optimizing the DFLAW flowsheet at Hanford with real waste. (authors)

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Radionuclide Inventory Calculations for the Calcined Solid Storage Facility Performance Assessment - 19292

Originally named the Idaho Chemical Processing Plant when it was constructed in the early 1950's, the Idaho Nuclear Technology and Engineering Center reprocessed SNF to recover highly enriched uranium and other nuclear-related products for the U.S. DOE. Reprocessing SNF generated millions of gallons of aqueous radioactive waste. Calcination was a process developed at the INL Site to convert liquid waste into a solid, granular form from 1963 to 2000. The calcination process produced a safer product for storage while reducing the volume of stored liquid waste by an average factor of seven. Calcine is currently stored in the Calcined Solids Storage Facility (CSSF). The CSSFs are often called 'bin sets' because each CSSF (CSSFs 1 through 6) contains several stainless-steel storage bins housed in a concrete vault. The CSSF is an interim waste storage facility. In the future, calcine will be retrieved, treated (if necessary), and sent to an off-Site facility for storage or disposal. To close the CSSF, the bulk of the calcine will be removed and stored or disposed of elsewhere. The bins and vaults will then be placed in a configuration that is protective of human health and the environment. For purposes of the performance assessment, the depth of the residual calcine at closure was assumed to be 5-cm in each bin. The 5-cm of residual calcine in each bin is assumed to be mixed in grout during closure. The activities of the residual radionuclides are the inventory used for the dose assessment calculations for the CSSF performance assessment. The objective of this analysis was to calculate the radionuclide activity in the residual calcine remaining in the CSSF after closure. The purpose of this paper is to present the approach used to calculate the radionuclide inventory of the calcine currently stored in CSSF and the residual calcine in the CSSF after closure using data from the calcined waste Historical Processing Model (HPM). Limited sampling and analysis of the calcine in the CSSF bins was conducted in the past. Future similar calcine sampling and analysis efforts would be costly, generate additional mixed waste, and may not provide the desired detailed calcine composition data. However, during the later years of calcining, an alternative to calcine sampling and analysis was developed and the information was captured in the HPM. The HPM uses multiple sources to provide definitive calcine volume, mass, and composition, including: compositions and volumes of the liquid wastes that were calcined, ORIGEN2 - and ORIGEN-ARP-based radioactivity calculations, calciner and CSSF operating data, and process knowledge of INTEC fuel and waste-processing systems. HPM calcine composition was used to develop an inventory of radionuclides for use in the performance assessment evaluation for closure of CSSF. There are 43 bins that contain calcine, with six to twelve segments in each bin for a total of 337 bin segments for which the HPM calculates the average calcine composition. The HPM calculates the composition of each of the nearly 7,000 batches of material (startup bed, nonradioactive material, and radioactive waste) charged to the calciners. The calculations use the CSSF dimensions, including the CSSF number of bins, heights, outside diameters, annular air space diameter (if necessary), and volume of calcine currently in the bins. For each of the CSSFs, for a 5-cm residual, the percent of total calcine was calculated along with the residual activity for the radionuclides. Residual activity was estimated for 148 different radionuclides. (authors)

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Engineering Challenges for the In Situ Disposal (ISD) of the 20 MWe Nuclear Power Demonstration (NPD) Reactor - 19407

The Nuclear Power Demonstration (NPD) reactor is the first CANDU{sup R} type power reactor to be decommissioned. This first of a kind project is to decommission the station using an In-Situ Disposal (ISD) strategy, and presents some engineering challenges to implement. Based on current technology and successful application of industry lessons learned, these challenges are being met. This paper presents a summary of those challenges, to recover from a long period of Storage With Surveillance, solve configuration management, assess the existing structure, determine the performance of the structure following seismic events during the disposal period, adapt the facility design to ISD and select a grout in-fill, and describe the regulatory and public interaction, to demonstrate the integrity of what will be a future disposal site. (authors)

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Waste Chemistry Envelope for the Determination of Pitting Corrosion Susceptibility - 19554

Liquid radioactive wastes at the Hanford and Savannah River sites are stored in large underground carbon steel tanks. The high-level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life (i.e., greater than 60 years). Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. Administrative programs are in place to mitigate potential corrosion mechanisms and thereby maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. Waste chemistry and temperature control are the means by which corrosion degradation of the waste tanks is minimized. Chemistry controls are based on in-situ coupon tests, waste tank experience, and laboratory testing. These requirements comprise the technical standards for corrosion control. The standards are utilized then as guidelines for operational procedures. The present waste chemistry controls for the Hanford DSTs were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits, which were developed at the Savannah River National Laboratory (SRNL) were coupled with corrosion testing results from Pacific Northwest National Laboratories (PNNL) and adopted for corrosion control of the double-shell tank (DST) waste at the Hanford Site. Since that time, the temperatures of the waste in the DSTs have decreased to less than 50 deg. C in most cases as the radioactive material has decayed. As a consequence, new chemistry controls were formulated for SCC in 2010. This testing however, did not address that potential for pitting in the supernatant and at the liquid-air interface. Indeed, initial tests indicated that the inhibitor requirements for SCC would be insufficient for the prevention of pitting. In addition to lower waste temperatures, future retrieval and process changes may impact the DST waste chemistry. Initial projections suggest that the waste chemistry of the DSTs may shift to a broader range of pH and higher aggressive anion concentrations (e.g., chloride, sulfate, etc.) than the wastes that are presently stored. Additionally, processes within the vitrification facility may result in streams that are returned to the tank farm for evaporation. These streams may also contain high concentrations of aggressive species. SRNL investigated the propensity for pitting of the carbon steel waste tanks at current and projected DST waste chemistries. A statistically designed series of tests provided a pitting factor that relates the ratio of the inhibitor species to the aggressive species and the probability of observing pitting. An acceptable pitting factor was determined and was utilized for the development of a new chemistry envelope. The new chemistry envelope was compared to the envelope that was established for SCC and was determined to be bounding. Validation of the new chemistry envelope was performed via model simulations and comparison with historical experimental results and approaches to establishing chemistry envelopes. The new chemistry envelope will eventually be part of an overall assessment of the condition of the tanks and any changes made to the overall structural integrity program for the Hanford DSTs. (authors)

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Minimizing Transuranic Waste Generated in Hot Cells at Oak Ridge National Laboratory Through Use of a Non-Destructive CO{sub 2} Pellet Cleaning System - 19577

Wastes generated in shielded hot cells containing materials contaminated with transuranic (TRU) waste present unique problems in waste disposal at Department of Energy (DOE) sites. Oak Ridge National Laboratory (ORNL) routinely processes materials contaminated with TRU waste from the High Flux Isotope Reactor (HFIR) and from other sources (both on and off-site) in their hot cell facilities, generating wastes that require eventual disposal. The waste streams generated by these processes are often high activity and exhibit high dose rates, driving them into the remote-handled (RH) category at the Waste Isolation Pilot Plant (WIPP). Reduction of the TRU waste footprint is desirable from a disposal cost perspective due to constraints with RH waste processing and a lack of space availability at the WIPP. Historically TRU waste generated at ORNL has been processed by the TRU Waste Processing Center (TWPC), which belongs to the Department of Energy's Office of Environmental Management (DOE-EM). Now that Oak Ridge legacy TRU waste is reaching its final stage, DOE-EM has turned over the mission of processing newly-generated TRU waste streams to the DOE Office of Science at ORNL. To reduce the volume of potential TRU waste occupying valuable space in the ORNL Hot Cells, a non-destructive CO{sub 2} pellet cleaning system will be used to capture particulate contaminants removed from the surfaces of waste inside the ORNL Non-Reactor Nuclear Facilities (NNFD) Hot Cells at the Irradiated Fuels Examination Laboratories (IFEL). The particulate will be segregated by particle size and captured for disposal later. This action will separate some of the TRU isotopes (particulate) from the larger waste volume (debris waste items), resulting in both lower waste disposal costs and reduced TRU waste volumes. This will be accomplished through a specialized cyclone separator. Secondary wastes are usually generated through other methods of decontamination, such as water jets and abrasive agents. These secondary wastes are absent with the CO{sub 2} pelletized system, due to sublimation of the dry ice upon impact. This is an important consideration with IFEL, a facility whose primary mission is to perform a series of R and D testing on nuclear fuel rods. Liquid is not a feasible method of decontamination as the IFEL's drains in the Hot Cells were plugged years ago due to Resource Conservation and Recovery Act (RCRA) regulations. Constraints at IFEL, such as criticality safety concerns with use of hydrogenous materials, are also alleviated with the non-destructive CO{sub 2} pellet cleaning system. These are a couple of the considerations why this method of decontamination is preferred. Due to concerns with oxidizers at WIPP grouting will be performed on the contaminated particulate produced during the trapping by the cyclone separation phase. This method will comply with mitigating oxidizing properties in TRU waste, required by the recently published Basis of Knowledge (BoK) document as part of Acceptable Knowledge (AK) by WIPP. (authors)

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Applying Technology and Experience to Ensure Safe In Situ Decommissioning - 19587

Permanent entombment of a radioactively contaminated facility as a decommissioning option (In Situ Decommissioning, ISD) has been completed for DOE facilities at the Savannah River Site (P and Reactors, HWCTR reactor), Hanford (U Canyon) and Idaho (CP-601/604),and is currently being considered for implementation at a limited number of selected DOE facilities as well as for the NPD and WR-1 reactors in Canada and various facilities in the United Kingdom. A general description of an ISD project encompasses an entombed facility which in some cases is limited to just the below-grade portion of a facility with the above-ground portion being decommissioned, packaged and disposed in a conventional manner. The entombed portions of the facility are of robust construction, generally of reinforced concrete exterior, that provides a migration barrier between internal contamination and the environment with significant internal void spaces backfilled or grouted. The scope of entombment may include ancillary equipment and structures and may contain radioactive and hazardous materials and contamination within the facility and waste imported from outside the facility. ISD is most often a permanent decommissioning end-state although in some instances (e.g. the U.K.) there may be a need to revisit the ISD facility after 300 years to effect traditional decommissioning. The defined completion (the end-state) of the decommissioned facility is therefore project-specific and in conformance with environmental approval processes and planned future activities. In all cases however, the final condition is passive, meaning that there are no requirements for ongoing operational systems or equipment within the decommissioned facility. ISD projects in the USA are presumed to be under indefinite institutional control of the U.S. Government. The regulatory framework in the U.S. is currently in place to provide assurance that the risk posed by an ISD facility is within regulatory acceptance criteria. (The framework in the U.K. and Canada is evolving towards that in the USA). Special emphasis is placed on the fact that an entombed facility is not considered a waste disposal facility; rather it is a decommissioning end-state option. ISD does not eliminate the need for effective and careful management of contaminated materials and structures, nor does it serve to abandon contaminated buildings in place. Further, ISD is safe and cost-effective for a relatively small number of facilities across the DOE Complex. This paper explores the current programmatic considerations and technical aspects of the ISD process and discuss the measures taken to ensure that an 'entombed' facility is safe for both site workers and the general population. (authors)

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Implementation of Increased Salt and Decreased Sludge Volumes in the Defense Waste Processing Facility - 19597

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). The disposition of high-level waste (HLW) into a glass waste form poured into canisters at the DWPF began in 1996. DWPF processed a sludge only flowsheet until 2008 when the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began treating the salt waste from the underground liquid storage tanks by striking with monosodium titanate (MST) to remove actinides, when necessary to meet downstream acceptance criteria, and removing the MST and entrained sludge solids (SS) by filtration before using an organic solvent to strip the cesium from the salt, creating two new concentrated HLW streams. This change caused the DWPF to begin processing under both a sludge-only and a coupled operations flowsheet where salt streams would only be added when available. This strategy allows for continuous glass canister production at DWPF. Recently, DWPF completed a 10-month melter replacement in December 2017 and the Concentrate, Storage, and Transfer Facility (CSTF) has completed repairs to the 3H Evaporator system. These facility outages have caused delays in preparing the next batch of sludge feed for DWPF. Concurrent with these facility outages, the Salt Waste Processing Facility (SWPF) has been undergoing system testing and commissioning, working toward startup. SWPF is designed to process salt waste in the same manner as ARP/MCU, but with much higher throughput volumes produced per unit time. In order to process the SWPF salt streams efficiently, DWPF will implement a coupled operation only flowsheet. Based on the facility outages mentioned above, efforts have been made to identify strategies to minimize sludge volumes and maximize salt volumes in the DWPF until the next sludge batch is ready for processing in DWPF. Savannah River Remediation (SRR) recently performed an evaluation that examines the effects of reducing the volume of sludge and maximizing the addition of the cesium-rich concentrated salt solution in each Sludge Receipt and Adjustment Tank (SRAT) batch in order to extend the existing sludge batch as long as possible until the next sludge batch is ready. These effects include a reduction of weight percent solids in the DWPF processing vessels, an increase in cycle time, a new canister heat generation value that remains within regulatory limits, and higher gamma source that remains within bulk facility gamma shielding. Balancing the above parameters resulted in a small reduction of sludge volume per SRAT batch under SWPF operation and maximizing the salt volumes for both ARP/MCU and SWPF operation. This slight reduction in sludge volume and increase in salt volumes allows SRR to mitigate some of the schedule impact incurred by the facility outages. This strategy was evaluated for both ARP/MCU and SWPF operation. This strategy has been successfully implemented under ARP/MCU operation using small step changes in the current process. This paper provides an overview of the evaluation performed for DWPF with ARP/MCU and DWPF with SWPF operations. (authors)

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Understanding and Controlling Cesium Loading on Crystalline Silicotitanate - 19601

Savannah River Remediation (SRR) contracted with Westinghouse Electric Corporation (WEC), Columbia Energy and Environmental Services (CEES), to build and deliver a modular ion exchange process that will be used to remove radioactive cesium from dissolved salt solution contained in Tank 10 at the Savannah River Site (SRS). The resulting decontaminated salt solution will be sent to Tank 11 and then to Tank 50 for final disposition in grout at the Saltstone facility. At the end of processing, the loaded ion exchange (IX) columns are drained, rinsed, and dried. The IX columns are removed from the modular unit and stored at Interim Safe Storage (ISS). This ion exchange process is called the Tank Closure Cesium Removal (TCCR) system. The TCCR system consists of several unit operations and process monitoring capability housed within three modular units. The first modular unit contains the process control computer, the second modular unit contains the main process equipment (two prefilters, four IX columns, and resin trap (post filtration)), and the third modular unit is the ventilation system. The main unit operations of the TCCR system are filtration, ion exchange, and ventilation. The ion exchange media utilized is a form of inorganic Crystalline Silicotitanate (CST), IONSIV{sup TM} R9120-B supplied by UOP Honeywell. CST has a high selectivity for cesium (Cs) and has some affinity for other Group IA alkali metals on the periodic chart (hydrogen, lithium, sodium, potassium, rubidium, and francium). CST has also been shown to have affinity for strontium and actinides. Cesium loading on the CST is dependent on many factors including chemical composition (cations and anions), temperature, and pH of the salt solution. The actual amount of Cs loaded onto an IX column is dependent on the factors above along with the salt solution volume fed to the IX column. Because of CST's high selectivity for Cs, the Cs-137 equilibrium loading can contribute enough decay heat to result in an ion exchange column boiling event under no flow conditions in the IX column. The boiling event under no flow is a safety concern that must be prevented from occurring. To prevent boiling of the column under no flow conditions, the Cs-137 equilibrium loading must be determined and understood. Several methodologies were evaluated for performing the Cs-137 equilibrium loading determination for CST. After review of available research and development (R and D), batch equilibrium contact test data, and Zheng Anthony Miller (ZAM) model predictions for the salt solution chemistry in Tank 10, a methodology for determining the Cs-137 equilibrium loading on CST was selected. The methodology selected involves an in-tank batch equilibrium contact test performed in duplicate that considers additional corrections to account for anticipated operational evolutions. The in-tank batch equilibrium contact test utilizes ∼0.1 grams of CST secured in a slotted sample vial. The slotted sample vial design allows for salt solution to be in contact with the CST. The slotted sample vials containing the CST are submerged in Tank 10 and exposed to the salt solution for ten days. After ten days, the samples are retrieved and sent to the Savannah River National Laboratory (SRNL) for analysis of the Cs-137 equilibrium loading. Additional corrections are then made to the Cs-137 results to account for anticipated operational evolutions. Anticipated operational evolutions include external additions (water/caustic) to Tank 10 that result in a change to its chemical composition, tank temperature effects, departures from the laboratory testing (mixed vs. stagnant tank conditions), and IX column flushes. A description of the methodologies considered, feasibility of methodology deployment, and the down select process (benefits vs. risks) is discussed in this paper. (authors)

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Challenges and Solutions in Operation and Lay-Up of the Modular Caustic- Side Solvent Extraction Unit - 20078

The Modular Caustic-Side Solvent Extraction (CSSX) Unit (MCU) was designed and constructed to provide interim salt waste processing so that the Liquid Waste Disposition Program (LWDP) could continue waste removal and tank closure until start-up of the Salt Waste Processing Facility (SWPF). MCU processes Clarified Salt Solution (CSS), or salt solution (SS) that has undergone actinide removal/filtration, to produce two waste streams: the cesium laden Strip Effluent (SE), which is incorporated into glass at the Defense Waste Processing Facility (DWPF), and the cesium depleted Decontaminated SS (DSS), which is incorporated into grout at the Saltstone Production Facility (SPF). MCU completed start-up in 2008 with an initial operating life of 3 years and design life of 5 years. Prior to reaching the end of the design life, multiple critical components were upgraded and/or replaced to mitigate risks associated with any delays in SWPF start-up. During the remainder of operation (until June 2019), MCU was challenged to maximize processing with minimal additional modifications to repair/improve aging infrastructure, while maintaining low levels of risk to workers and the environment. Notable challenges during the extended operating life of the facility involved both mechanical and operational issues. Two of the most significant issues were the increased frequency of Process Vessel Ventilation (PVV) High Efficiency Particulate Air (HEPA) filter change-outs and the biofouling of the SE coalescer (SEC). The major impacts of both issues were that replacement of these components 1) resulted in high exposure to workers, and 2) required substantial downtime, thus hindering achievement of processing goals. PVV HEPA filter replacements are required based on dose rate and differential pressure (dP) limits. After introducing a higher-curie feed to MCU, the operating time between change outs was reduced by approximately half. SEC replacements are required based on dP limits. After restarting from an extended outage, the operating life of the SEC was significantly reduced by more than half due to biofouling. As a result of extensive troubleshooting and process improvements, the facility was able to recover performance and extend the operating life of the PVV HEPA filters and the SEC media. Through troubleshooting both of these issues, the facility was also able to learn from prior experience and adapt in order to minimize down time and maintain throughput. Although the facility took advantage of short periods of downtime to perform nonintrusive troubleshooting and minor corrective/preventative maintenance activities, longer outages were periodically required for corrective maintenance involving process cell entry. During these outages, remote tools and specialized shield plates were employed to minimize dose to workers. Mock-ups of non-routine maintenance activities were critical to early identification of potential issues and improvements, which shortened outage duration. Consistent with the facility's operational strategy, evolutions needed for MCU layup were identified and sequenced to ensure safe conditions within the facility, meet pre-defined criteria for lay-up configuration, and protect workers, while minimizing the impact to SWPF integration. De-inventory and flushing of process areas was prioritized so that the total residual inventory was reduced as efficiently as possible. This paper discusses lessons learned and best practices from troubleshooting unique issues, performing complex maintenance activities, and implementing a layup strategy that best supports the overall system mission. (authors)

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Improving PCM Waste Management Through New Approaches to Screening for LLW Treatment and Disposal - 20145

A major step forward in the management of Plutonium Contaminated Material (PCM) Waste has been taken at the Sellafield site with the introduction of new capabilities to enable wastes previously conservatively designated as PCM Intermediate Level Waste (ILW) to be re-classified for treatment as Low Level Waste (LLW). The Sellafield site, located in the North West of England, within the United Kingdom (UK), has operated plutonium processing plants and waste facilities for almost 70 years. During this time management of the raw PCM waste produced has presented a variety of challenges. The waste is currently processed by high force waste compaction in the Waste Treatment Complex (WTC). The WTC converts 200 litre drums of raw PCM waste into a compacted, grout encapsulated monolith form within stainless steel containment, ready for emplacement within the UK's Geological Disposal Facility (GDF) when it becomes available. To help progress some of the wastes waiting for treatment, recent improvements have been made with the introduction of new waste assay instrumentation which enables the identification of a 'candidate' waste potentially suitable for disposal via the LLW treatment route. Opening a new LLW route for previously designated PCM has been a major advance in PCM waste management policy. It has required a range of technical assessment arrangements to be developed, tested and commissioned and involved the engagement of a wide range of stakeholders. In particular, very detailed technical analysis methods have been developed to ensure that there is a robust and conservative approach to the final sentencing of waste. Following completion of commissioning tests, the first 86 m{sup 3} of previously designated PCM waste was successfully treated through the LLW route and consigned to LLW disposal in June 2019. The direct benefit of sentencing waste to the correct waste route (LLW) is the significant cost savings across the complete range of waste management activities. This work is also strategically important for demonstrating to a range of stakeholders including workforce and regulators, that waste previously designated as PCM can be re-classified in a systematic and justifiable way without increasing risk to personnel or the wider environment. By ensuring that waste is being managed appropriately it will be possible to carry out more remediation work safely, in less time, and at a lower cost. (authors)

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A Decade of Innovative Approaches to Treating and Dispositioning Radioactive Waste at the Savannah River Site - 20620

This year marks the 70. anniversary of the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina. SRS is a key industrial complex responsible for environmental stewardship, environmental cleanup, waste management, and disposition of nuclear materials. SRS also continues to have a role in critical defense-related activities and the reprocessing of used reactor fuel. The SRS encompasses 803 square kilometers (310 square miles) in parts of Aiken, Barnwell, and Allendale counties. The Liquid Waste Mission starts with the safe receipt and storage of radioactive liquid waste in the waste tanks, which principally includes Cold War legacy waste, but also waste from support of National Aeronautics Space Administration missions, medical isotope production, and research activities. This 133,000 cubic meters (m{sup 3}) (35 million gallons [Mgal]) of high-level radioactive waste (HLW) is currently held in 43 large underground waste tanks. The capacity of each of these tanks range from 2,800 m{sup 3} to 5,000 m{sup 3} (0.75 to 1.33 Mgal) and were placed in operation between 1954 and 1986. In July 2009, the DOE awarded a stand-alone contract to execute the Liquid Waste Mission at SRS. The Liquid Waste Contractor selected was Savannah River Remediation LLC1 (SRR), which is responsible for receipt, storage, retrieval and treatment of all HLW, disposal of the decontaminated low-activity waste fraction, as well as operationally closing cleaned HLW tanks. Since July 2009, SRR has successfully grouted and operationally closed six large underground tanks, poured 1,476 canisters of vitrified HLW (a stable glass waste form), treated approximately 34,200 m{sup 3} (9.0 Mgal) of HLW, and dispositioned over 41,200 m{sup 3} (10.9 Mgal) of decontaminated salt solution as low-level waste (LLW) into the Saltstone Disposal Units (SDUs) being constructed by SRR on site. The innovative work performed by SRR in the LLW portion of its mission has earned it the Richard S. Hodes Award. The work conducted by SRR is the same work championed by the late Richard S. Hodes, a respected physician, statesmen, and Chairman of the Southeast Compact Commission for Low-Level Radioactive Waste Management. (authors)

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Use of Unmanned Aerial Systems for Post-closure Waste Site Maintenance and Surveillance - 20143

The P- and R-Area Reactor buildings, located at SRS near Aiken, SC, were in-situ decommissioned by grouting below-grade portions of the buildings and demolishing some above-grade structures to grade level. Other reactor building structures were left above-grade and sealed to prevent human or animal access. Because the above grade-structures are expected to continue in their present state for hundreds of years, the condition of the building roofs is critical to mitigate rainwater intrusion. For this reason, building roof areas were strengthened with high strength concrete to ensure the long-term integrity of the roof structure. Periodic inspections of the roof structures are required to ensure that the roofs are functioning properly and to identify damage areas. The traditional method of inspecting closed reactor buildings requires the use of a helicopter and photographer to generate photographs and videos for review. This method proved sufficient but lacked the resolution and clarity that is required for a thorough inspection of the building structure. Fortunately, the Savannah River National Laboratory (SRNL) has established a Small Unmanned Aircraft System (sUAS) program using commercially available and custom-built remote-controlled aircraft. EC and ACP contacted the SRNL program to evaluate whether the sUAS technology could be employed for periodic reactor building inspections. The sUAS can fly within a few meters of the reactor buildings and hover, allowing for a more thorough aerial inspection. The initial sUAS inspection of the P-Area Reactor building was completed in February 2018 and at the R-Area Reactor building in August 2018 and were successful in providing higher resolution photos and videos. The sUAS inspections also revealed that vegetation had begun to grow on the roofs that could potentially damage the structure integrity. SRNL partnered with Virginia Polytechnic Institute and State University to build a custom heavy lift sUAS that could be equipped with herbicides to remotely treat the vegetative growth. Herbicides were successfully applied to the R-Area Reactor building roof using the custom sUAS in September 2018. The use of unmanned aircraft systems at SRS to perform aerial inspections and herbicide treatment in otherwise inaccessible areas has proven to be an efficient and cost-effective technology that provides high value, increases knowledge of facility conditions, and provides for early detection of damage. Periodic inspections of the reactor building roofs using sUAS technology are performed safely, efficiently, and at a significant cost and schedule savings while reducing emissions, noise, and fossil fuel use. The use of sUAS equipment for building inspections and herbicide application at SRS is unique within the DOE complex. The goal of SRS is to apply this remote technology to other waste unit operations and maintenance activities in the future. (authors)

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SRS Strategy for Tank 3 Salt Dissolution - Performance and Lessons Learned - 20441

The Savannah River Site (SRS) Tank Farms have 51 underground waste tanks used to store and process liquid nuclear waste materials. There are 4 different tank types, ranging in capacity from 2,840,000 to 4,920,000 L (750,000 to 1,300,000 Gal). Twenty-four of the tanks are older style and do not meet full secondary containment standards. The older style tanks are the initial focus of waste removal efforts for tank closure at SRS. Eight of these twenty-four tanks have completed waste removal and are filled with grout. Prior to salt dissolution, Tank 3 was a dry salt tank that contained 5.06 m (199.3 in) of salt and sludge waste. Additional salt waste was present on cooling coils above the salt layer up to approximately 5.59 m (220 in). Three mixing eductors were installed in Tank 3 to aid in dissolving salt waste in three tank riser access ports. A transfer pump was installed, and the transfer pump suction was located 25.4 cm (10 in) from the tank bottom. Well water was added to the tank through a downcomer until the dry bulk salt was covered with liquid. During the initial fill of Tank 3, approximately 242,000 L (64,000 Gal) of rain water were added from periodic F-Tank Farm (FTF) Catch Tank additions and approximately 17,000 L (4,500 Gal) of well water were added. Following liquid additions to cover the dry bulk salt, well water was added through the three mixing eductors in batches during each stage. Additionally, during the salt dissolution campaigns, the FTF Catch Tank was utilized to add rain water through a downcomer in the center tank riser access port as needed for volume relief in the FTF Catch Tank. Following liquid additions, the water was recirculated (internal to the tank) using the transfer pump, and a sample was pulled to confirm the target specific gravity (SpG) of the dissolved salt solution was achieved. The dissolved salt solution was then transferred to the receipt tank (Tank 7), and the mixing eductors were lowered as close to the new bulk salt layer as possible, to support subsequent dissolution campaigns. While the mixing jets were able to dissolve salt successfully in Tank 3, they did not do so in a completely uniform manner. Throughout dissolution, mounds were discovered under Riser 1, Riser 2, and the Center Riser. The mixing jets dissolved the salt around the edges of the tank well but were not as effective toward the center of the tank. FTF Catch Tank additions through the center tank riser access port downcomer were required to impact the mound under the center tank riser. FTF Catch Tank additions were sporadic as they were dependent on rainfall in the area. Additionally, all mixing jets were rarely able to be lowered to the same elevation, indicating some mounding in the bulk salt layer. Indexing of the mixing jets was also utilized to impact the salt mounds. Overall, Tank 3 salt dissolution was successful as approximately 1,476,000 L (390,000 Gal) of dissolved salt solution was transferred to Tank 7 over six stages. Slightly less than 852,000 L (225,000 Gal) of water were added to Tank 3. The original material balance prediction estimated that the bulk salt level in Tank 3 would be 3.88 m (152.8 in) after six salt dissolution stages. After six stages the actual bulk salt level was 3.38 m (133 in). This paper will discuss the salt removal strategy, each salt dissolution stage, and lessons learned for future salt dissolution. (authors)

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Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470

The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)

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The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

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