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Design of an Evapotranspiration Surface Barrier for the Subsurface Disposal Area at INL - 19220

At the INL Site, the US DOE is planning to construct one of its largest CERCLA landfill covers. An evapotranspiration surface barrier that will be approximately 65 ha (160 acres) will cover the 39-ha (97-acre) Subsurface Disposal Area at the INL to isolate legacy mixed waste disposed of prior to 1988, low-level radioactive waste disposed of after 1988, and debris from demolition of structures being used to process some of the legacy mixed waste. The evapotranspiration surface barrier will limit percolation to less than 1 cm (0.39 in.) per year under typical precipitation conditions; provide a minimum 3-m (10-ft)-thick cover over all waste to minimize the potential for plant, animal, and inadvertent human intrusion; support growth of native vegetation; and visually blend in with the surrounding desert landscape. A robust design will allow the evapotranspiration surface barrier to resist erosion and perform for a minimum of 1,000 years. The 1-cm (0.39-in.)-per-year and 1,000-year performance objectives were established by a CERCLA record of decision and its implementing remedial design work plan, and these performance objectives are consistent with a performance assessment conducted under US DOE Order 435.1, 'Radioactive Waste Management.' (authors)

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Improving Site-Specific Radiological Performance Assessments - 13431

An improved approach is presented for conducting complete and defensible radiological site-specific performance assessments (PAs) to support radioactive waste disposal decisions. The basic tenets of PA were initiated some thirty years ago, focusing on geologic disposals and evaluating compliance with regulations. Some of these regulations were inherently probabilistic (i.e., addressing uncertainty in a quantitative fashion), such as the containment requirements of the U.S. Environmental Protection Agency's (EPA's) 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, Chap. 191.13 [1]. Methods of analysis were developed to meet those requirements, but at their core early PAs used 'conservative' parameter values and modeling approaches. This limited the utility of such PAs to compliance evaluation, and did little to inform decisions about optimizing disposal, closure and long-term monitoring and maintenance, or, in general, maintaining doses 'as low as reasonably achievable' (ALARA). This basic approach to PA development in the United States was employed essentially unchanged through the end of the 20. century, principally by the U.S. Department of Energy (DOE). Performance assessments developed in support of private radioactive waste disposal operations, regulated by the U.S. Nuclear Regulatory Commission (NRC) and its agreement states, were typically not as sophisticated. Discussion of new approaches to PA is timely, since at the time of this writing, the DOE is in the midst of revising its Order 435.1, Radioactive Waste Management [2], and the NRC is revising 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste [3]. Over the previous decade, theoretical developments and improved computational technology have provided the foundation for integrating decision analysis (DA) concepts and objective-focused thinking, plus a Bayesian approach to probabilistic modeling and risk analysis, to guide improvements in PA. This decision-making approach, [4, 5, 6] provides a transparent formal framework for using a value- or objective-focused approach to decision-making. DA, as an analytical means to implement structured decision making, provides a context for both understanding how uncertainty affects decisions and for targeting uncertainty reduction. The proposed DA approach improves defensibility and transparency of decision-making. The DA approach is fully consistent with the need to perform realistic modeling (rather than conservative modeling), including evaluation of site-specific factors. Instead of using generic stylized scenarios for radionuclide fate and transport and for human exposures to radionuclides, site-specific scenarios better represent the advantages and disadvantages of alternative disposal sites or engineered designs, thus clarifying their differences as well as providing a sound basis for evaluation of site performance. The full DA approach to PA is described, from explicitly incorporating societal values through stakeholder involvement to model building. Model building involves scoping by considering features, events, processes, and exposure scenarios (FEPSs), development of a conceptual site model (CSM), translation into numerical models and subsequent computation, and model evaluation. These are implemented in a cycle of uncertainty analysis, sensitivity analysis and value of information analysis so that uncertainty can be reduced until sufficient confidence is gained in the decisions to be made. This includes the traditional focus on hydrogeological processes, but also places emphasis on other FEPSs such as biotically-induced transport and human exposure phenomena. The significance of human exposure scenarios is emphasized by modifying the traditional acronym 'FEPs' to include them, hence 'FEPSs'. The radioactive waste community is also recognizing that disposal sites are to be considered a national (or even global) resource. As such, there is a pressing need to optimize their utility within the constraints of protecting human health and the environment. Failing to do so will result in the need for additional sites or options for storing radioactive waste temporarily, assuming a continued need for radioactive waste disposal. Optimization should be performed using DA, including economic analysis, invoked if necessary through the ALARA process. The economic analysis must recognize the cost of implementation (disposal design, closure, maintenance, etc.), and intra- and inter-generational equity in order to ensure that the best possible radioactive waste management decisions are made for the protection of both current and future generations. In most cases this requires consideration of population or collective risk. (authors)

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Maximum Potential Hydrogen Gas Retention in the sRF Resin Ion Exchange Column for the LAWPS Process

The Low-Activity Waste Pretreatment System (LAWPS) is being developed to provide treated supernatant liquid from the Hanford tank farms directly to the Low-Activity Waste (LAW) Vitrification Facility at the Hanford Tank Waste Treatment and Immobilization Plant. The design and development of the LAWPS is being conducted by Washington River Protection Solutions, LLC. A key process in LAWPS is the removal of radioactive Cs in ion exchange (IX) columns filled with spherical resorcinol-formaldehyde (sRF) resin. One accident scenario being evaluated is the loss of liquid flow through the sRF resin bed after it has been loaded with radioactive Cs and hydrogen gas is being generated by radiolysis. In normal operations, the generated hydrogen is expected to remain dissolved in the liquid and be continuously removed by liquid flow. For an accident scenario with a loss of flow, hydrogen gas can be retained within the IX column both in the sRF resin and below the bottom screen that supports the resin within the column. The purpose of this report is to summarize calculations that estimate the upper-bound volume of hydrogen gas that can be retained in the column and potentially be released to the headspace of the IX column or to process equipment connected to the IX column and, thus, pose a flammability hazard.

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Nuclear aircraft carrier inactivation - Execution of waste management initiatives - 15565

After 6 years of planning efforts, in June 2013, Newport News Shipbuilding (NNS) began the execution phase of the inactivation of the USS Enterprise. The USS Enterprise decommissioning involves implementation of a new approach to defueling aircraft carriers, utilizing new processes for defueling and shipment of spent nuclear fuel and waste. During planning for this complex endeavor, NNS developed several process improvement initiatives including streamlining waste management processes and technology insertion using modeling and simulation (M and S) capabilities. One of the new approaches involves the establishment of a Waste Packaging Area (WPA) onboard the ship. The process assists with maximizing waste loading and shipments directly from the ship. Another waste management initiative includes the NNS/ Waste Vendor teaming effort to facilitate more efficient loading and shipment of inactivation waste. With execution underway, these initiatives are being put to the test to see if they can withstand the rigors of day-to-day production. NNS has implemented the use of the Modeling and Simulation (M and S) technology. During the planning phase, the M and S 3D Spatial Arrangement Tool was utilized to determine the optimal equipment layout within the new M-290 facility used for spent fuel packaging and ship-out and to determine the appropriate shipping container placement on the ship to maximize waste loading. The M and S technology proved to be a beneficial application resulting in continued utilization through the execution phase of the carrier inactivation. NNS is currently expanding the M-290 Facility Model to provide a visual representation of the facility, its processes, resources, and variable parameters that can be utilized as a decision support tool through simulation of the spent fuel packaging and ship-out process through the facility's life-cycle. NNS is also expanding the application of the waste management model to effectively predict waste forecast. The simulation model is linked with the defueling schedule which captures the operational elements of the waste flow process. The model extrapolates the data to provide waste generation output to assess the most efficient and cost effective path for radiological waste shipments offsite. The ability to conduct waste forecasting allows streamlining of the waste management process in support of complex defueling operations. (authors)

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Scientific Investigations on the Chemical Reactivity of Los Alamos Remediated Nitrate Salt Wastes - 16540

The February 14, 2014 release of radioactivity at the Waste Isolation Pilot Plant (WIPP) was the result of runaway thermal reactions that occurred in a TRU waste drum containing nitrate salts remediated using an organic absorbent. Experimental and modeling studies performed at LANL indicate that mixtures of metal nitrate salts (oxidizer) with sWheat Scoop{sup R} organic kitty litter (fuel) create the potential for exothermic chemical reactions. The use of sWheat{sup R} absorbent in the processing of nitrate salt wastes can be pinpointed as the critical processing decision that led to the failure of the breached drum in the WIPP repository. The most plausible scenario is that a production of heat, either from low-level chemical reactions or the growth of natural microbes, in concert with mixed metal nitrate salts, bismuth lined glovebox gloves and/or lead nitrates when combined with the sWheat{sup R} organic kitty litter, generated a series of exothermic reactions that heated and pressurized the drum resulting in the venting of high-temperature gases and radioactive material into the room. Observations and modeling studies indicate that the probability of thermal runaway in another drum has decreased with time since the WIPP event. However, the complexity of the mixtures, ambiguity in procedures such as those used for neutralization, the heterogeneity of the drum contents, and the difficulty of sampling leads to an irreducible level of uncertainty that mandates the exercise of caution in managing RNS wastes. The situation also requires that this waste stream continue to be stored in a temperature-controlled environment and monitored until the waste can be remediated to remove the hazard permanently. We recommend that proposals to treat nitrate salts with organic absorbents of any kind be accompanied by experimental studies to confirm that similarly sensitive mixtures are not inadvertently created over the range of conditions (mixture quantities, moisture content, etc.) anticipated during processing and storage. (authors)

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Environmental cleanup of the East Tennessee Technology Park - Year three - A status report - 15522

On August 1, 2011, URS - CH2M Oak Ridge LLC (UCOR) began its five-year, $1.4 billion cleanup of the East Tennessee Technology Park (ETTP), the former Oak Ridge Gaseous Diffusion Plant, located on the U.S. Department of Energy's (DOE) Oak Ridge Reservation (Reservation) in Tennessee. Under the contract, the base period scope of work to be completed by 2016 includes demolition of both the K-25 and K-31 gaseous diffusion buildings and all demolition waste dispositioned, initiation of demolition on the remaining K-27 gaseous diffusion building, safe storage of transuranic (TRU) waste for eventual transfer to the Transuranic Waste Processing Center, previously designated -No-Path-To-Disposition Waste- dispositioned to the extent possible, and completion of design and engineering of a mercury treatment system to be installed at the Y-12 National Security Complex (Y-12). UCOR will have also managed DOE Oak Ridge Office of Environmental Management (OREM)-owned facilities at ETTP, Oak Ridge National Laboratory (ORNL), and Y-12 in a safe and cost-effective manner. Since assuming responsibilities as the ETTP and Oak Ridge Reservation cleanup contractor, UCOR has completed its life-cycle Performance Measurement Baseline; received its Earned Value Management System (EVMS) certification; completed the deactivation and demolition (D and D) of the 44-acre K-25 gaseous diffusion building; removed and disposed of a contaminated underground storage tank (Tank W-1A) at ORNL; dispositioned mercury-contaminated storage tanks at Y-12; completed closure of the Toxic Substances Control Act Incinerator; disposed of more than 158,000 cubic yards of cleanup waste while managing the on-site Environmental Management Waste Management Facility (EMWMF); and conducted operations, surveillance, and maintenance activities at 180 DOE EM facilities at ETTP, ORNL, and Y-12 in a safe and cost-effective manner. In 2014, UCOR completed the removal and disposal of all of the K-25 demolition debris; completed the removal of asbestos-laden transite siding on the K-31 gaseous diffusion building and began demolition of that 17-acre building; advanced the characterization and deactivation of the K-27 gaseous diffusion building, readying the building for the start of demolition in 2015; advanced the design and engineering of a Mercury Treatment Facility to be located at Y-12; continued the cost-effective and safe disposition of all cleanup wastes; and readied ETTP for Voluntary Protection Program Star Status Assessment in 2015. Project performance as of December 31, 2014, continued to be excellent: - Cost Performance Index - 1.12 - Schedule Performance Index - 1.02 At the same time, since safety is the foundation of all cleanup work, UCOR's safety record goes hand-in-hand with its excellent project performance. Through calendar year 2014, UCOR's recordable injury rate was 1.04 occurrences per 200,000 work hours, and the lost workday case rate was 0.52. This performance continues to be due, in large part, to the people and processes URS and CH2M HILL, the parent companies of UCOR, brought to the project. Key enablers included: - Selected and deployed experienced staff in key leadership positions throughout the organization, continually re-energizing the organization with new leaders - Implemented previously successful safety programs and initiatives from other DOE sites, such as DOE's Voluntary Protection Program - Brought fresh thinking and innovation to tackling tough project issues, such as the deactivation of the Technetium-99-contaminated areas of K-27, applying lessons learned from the D and D of K-25; and implementing a -waste factory- approach to cleanup - Maintained continuing alignment between UCOR's contract with DOE and its Performance Measurement Baseline - rapidly addressing contract changes when they arose - Maintained a readiness both from a process as well as experienced personnel perspective to support DOE OREM in performing other significant Oak Ridge Reservation cleanup activities. (authors)

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Enterprise SRS: Leveraging Ongoing Operations to Advance Nuclear Fuel Cycle Programs - 12579

The international leadership in nuclear technology development and deployment long held by the United States has eroded due to the lack of clear national strategies for advanced reactor fuel cycle concepts and for nuclear materials management, as well as to the recent policy decision that halts work on the nuclear fuel repository at Yucca Mountain. Although no national consensus on strategy has yet been reached, a number of recent high-profile reviews and workshops have clearly highlighted a national need for robust research, development and deployment (RD and D) programs in key areas of nuclear technology, especially nuclear separations science and engineering. Collectively, these reviews and workshops provide a picture of the nuclear separations mission needs for three major program offices: Department of Energy Office of-Environmental Management), DOE Office of Nuclear Energy), and the National Nuclear Security Administration (NNSA). While the individual program needs differ significantly in detail and timing, they share common needs in two critical areas of RD and D: - The need for access to and use of multi-purpose engineering-scale demonstration test facilities that can support testing with radioactive material, and - The need for collaborative research enterprises that encompass government research organizations (i.e., national laboratories), commercial industry and the academic community. Such collaborative enterprises effectively integrate theory and modeling with the actual experimental work at all scales, as well as strengthen the technical foundation for research in critical areas. The arguments for engineering-scale collaborative research facilities are compelling. Processing history has shown that test programs and demonstrations conducted with actual nuclear materials are essential to program success. It is widely recognized, however, that such facilities are expensive to build and maintain; creating an imposing, if not prohibitive, financial burden on an individual sponsoring office. Given that reality, success for the current and future nuclear separations missions is dependent on a concerted effort to develop new, creative, approaches that leverage existing facilities in a manner that supports both near- and long-term needs of national programs. As a result of this situation, the Savannah River National Laboratory (SRNL) organized the 'Nuclear Separations User Facility Strategy Session' in Washington, D.C. on July 29, 2011. This workshop brought together key stakeholders from DOE and the private sector to develop a strategy for using engineering-scale nuclear materials processing facilities to advance our nation's nuclear separations research needs. In particular, the meeting focused on recommending how these engineering-scale demonstration facilities, like the Savannah River Site H-Canyon, can be connected with smaller 'bench-scale' research activities to form a seamless approach that integrates across the continuum of RD and D of advanced separations technologies. Coming out of this workshop, a new vision has been developed for a collaborative research facility model that centers on H-Canyon. Unique to this approach is the fact that H-Canyon will continue to accomplish DOE's critical nuclear material processing missions, while simultaneously serving as an RD and D resource for the scientific and technical portions of the nuclear separations community. This paper describes the planned operations for H-Canyon in FY2012 and beyond and discusses how these operations fit within the context of a collaborative research facility model and support the ongoing fuel cycle research and development programs of the DOE. (authors)

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Generic FEPs Catalogue and Salt Knowledge Archive

This report describes the development of a comprehensive catalogue of generic features, events, and processes (FEPs) that are potentially important for the post-closure performance of a repository for high-level radioactive waste (HLW) and spent nuclear fuel (SNF) in salt (halite) host rock. The FEPs and other supporting information have been entered into a “SaltFEP” Database. The generic salt repository FEPs include consideration of relevant FEPs from a number of U.S., Dutch, German, and international FEP lists and should be a suitable starting point for any repository program in salt host rock. The salt FEP catalogue and database employ a FEP classification matrix approach that is based on the concept that a FEP is typically a process or event acting upon or within a feature. The FEP matrix provides a two-dimensional structure consisting of a Features/Components axis that defines the “rows” and a Processes/Events axis that defines the “columns” of the matrix. The design of the FEP classification matrix is consistent with repository performance assessment – the Features/Components axis is organized vertically to generally correspond to the direction of potential radionuclide migration (from the waste to the biosphere) and the Processes/Events axis is designed to represent the common two-way couplings between thermal processes and other processes (such as thermal-mechanical or thermal-hydrologic processes). Related FEPs can be easily identified – related FEPs will typically be grouped in a single matrix cell or aligned along a common row (Feature/Component) or column (Process/Event). The online SaltFEP database can be downloaded from www.saltfep.org. It contains the FEP matrix, the FEPs, and the associated processes for each FEP. It provides a starting point to create and document site-specific individual FEPs. Furthermore, the FEP matrix is connected to the Salt Knowledge Archive (SKA), a database of about 20,000 references and documents representing the historical knowledge on radioactive disposal in salt. This work is the result of an ongoing collaboration between researchers in the U.S., the Netherlands, and Germany, and supports the NEA Salt Club Mandate. It builds upon prior work which is documented.

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Siting Radioactive Waste Management Facilities: Successes and Failures - 16392

Our assessment in 2015 of 39 attempts to site nuclear waste facilities in democratic countries over the past 40 years identified common inherent threads of both success and failure. In the successes, including those efforts still ongoing, the common threads include a) providing the host communities the opportunity to withdraw at any point, b) the physical presence of a significant project team in the host community, and c) the establishment of credibility and trust in the project team, the regulator and, where possible, a neutral commenting party. In the failures, two dominant causes are present. The first is the failure to properly address the legitimate concerns of indigenous people. This has caused failures in the US and Australia. The second and more common cause is lack of political consensus amongst the host community. In nearly every case when the facility proponent selects the site and attempts to gain host approval, the effort has failed. Within the failed cases, two topics appear to be common. The first is that the definition of 'host community' is initially too narrow by the proponents. In many situations, the very local community is supportive of radioactive waste projects but the larger community is not. The second factor is a failure on the part of the proponent to accurately communicate the risks associated with the project, driven significantly by differences in the manner in which the general public and experts process risk communications. (authors)

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Extraction of lead from waste CRT funnel glass by generating lead sulfide – An approach for electronic waste management

Highlights: • Sodium carbonate and free silica in the glass react to generate sodium silicate and carbon dioxide. • Sodium sulfide and lead silicate group of glass began to react to lead sulfide and sodium silicate. • Lead sulfide got together in the bottom of the crucible under the action of gravity. • The generated lead sulfide was separated and collected in the bottom of the crucible. - Abstract: Waste cathode ray tube (CRT) funnel glass is the key and difficult points in waste electrical and electronic equipment (WEEE) disposal. In this paper, a novel and effective process for the detoxification and reutilization of waste CRT funnel glass was developed by generating lead sulfide precipitate via a high-temperature melting process. The central function in this process was the generation of lead sulfide, which gathered at the bottom of the crucible and was then separated from the slag. Sodium carbonate was used as a flux and reaction agent, and sodium sulfide was used as a precipitating agent. The experimental results revealed that the lead sulfide recovery rate initially increased with an increase in the amount of added sodium carbonate, the amount of sodium sulfide, the temperature, and the holding time and then reached an equilibrium value. The maximum lead sulfide recovery rate was approximately 93%, at the optimum sodium carbonate level, sodium sulfide level, temperature, and holding time of 25%, 8%, 1200 °C, and 2 h, respectively. The glass slag can be made into sodium and potassium silicate by hydrolysis in an environmental and economical process.

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Classification methodology for tritiated waste requiring interim storage

Fusion machines like the ITER experimental research facility will use tritium as fuel. Therefore, most of the solid radioactive waste will result not only from activation by 14 MeV neutrons, but also from contamination by tritium. As a consequence, optimizing the treatment process for waste containing tritium (tritiated waste) is a major challenge. This paper summarizes the studies conducted in France within the framework of the French national plan for the management of radioactive materials and waste. The paper recommends a reference program for managing this waste based on its sorting, treatment and packaging by the producer. It also recommends setting up a 50-year temporary storage facility to allow for tritium decay and designing future disposal facilities using tritiated radwaste characteristics as input data. This paper first describes this waste program and then details an optimized classification methodology which takes into account tritium decay over a 50-year storage period. The paper also describes a specific application for purely tritiated waste and discusses the set-up expected to be implemented for ITER decommissioning waste (current assumption). Comparison between this optimized approach and other viable detritiation techniques will be drawn. (authors)

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Processing and disposition of remote-handled transuranic liquid waste generated at Oak Ridge National Laboratory - 15113

Oak Ridge National Laboratory (ORNL) produces transuranic (TRU) isotopes for the US Department of Energy's (US DoE's) heavy-element research program and other programs for medical and industrial applications. As a result, the research facilities generate remote-handled TRU liquid low-level waste (LLLW). This LLLW is presently collected and mixed with larger volumes of legacy and environmental management waste streams for future treatment. Once the legacy waste is disposed of as part of the US DOE Environmental Management program, the large-scale waste treatment processes will be decommissioned, and new capabilities will be needed to characterize, treat, package, and dispose of the newly generated LLLW at ORNL. Various approaches are being considered for addressing treatment and disposal of the future LLLW streams. A variety of processing methods and final waste forms is under consideration. The impacts of the chemical and radionuclide compositions of the waste stream are being evaluated. (authors)

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HEPA Filter Disposal Write-Up 10/19/16

Process knowledge (PK) collection on HEPA filters is handled via the same process as other waste streams at LLNL. The Field technician or Characterization point of contact creates an information gathering document (IGD) in the IGD database, with input provided from the generator, and submits it for electronic approval. This document is essentially a waste generation profile, detailing the physical, chemical as well as radiological characteristics, and hazards, of a waste stream. It will typically contain a general, but sometimes detailed, description of the work processes which generated the waste. It will contain PK as well as radiological and industrial hygiene analytical swipe results, and any other analytical or other supporting knowledge related to characterization. The IGD goes through an electronic approval process to formalize the characterization and to ensure the waste has an appropriate disposal path. The waste generator is responsible for providing initial process knowledge information, and approves the IGD before it routed to chemical and radiological waste characterization professionals. This is the standard characterization process for LLNL-generated HEPA Filters.

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EBR-II Driver Fuel Cladding Hull Analytical Data

Sharing EBR-II driver fuel cladding hull data with General Atomics under NDA No. 20NDA155. Data share is needed to advance discussions regarding process technologies to decontaminate separated cladding hulls

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Modeling of Glycolate Destruction in the Recycle Collection Tank

The Savannah River Site’s DWPF is being upgraded with the introduction of the NG flowsheet. Glycolic acid has been shown to be a superior alternative to formic acid for sludge processing. The new flowsheet improves or maintains necessary parameters such as 1) reduction of mercury, 2) adjustment of feed rheology, 3) pH stability, and 4) adjustment of melter oxidation/reduction potential. Further, the use of glycolic acid virtually eliminates the potential for catalytic hydrogen generation in DWPF processing DWPF process condensates are collected and returned to the Savannah River Site (SRS) CSTF. The RCT collects off-gas condensate during chemical processing, vitrification, and other unit operations performed in DWPF and is the singular return vessel delivering recycle effluent back to CSTF. Each batch of recycle will have a small amount of glycolate from chemical processing and melter off-gas condensates. To avoid potential flammability issues due to thermolysis of glycolate in the CSTF, Savannah River National Laboratory (SRNL) provided to Savannah River Remediation (SRR) at their request a Task Technical and Quality Assurance Plan (TTQAP) to quantify and mitigate glycolate returns via DWPF’s recycle stream. The request included testing of a process to oxidize glycolate and other organic species that are responsible for hydrogen generation from thermolysis. Following that work SRR provided a Task Technical Request (TTR) that requested process modeling. In 2021 a TTQAP was issued to cover the modeling work. Modeling draws data from laboratory scale studies using chemical simulants and radioactive waste samples. Chemical kinetic modeling was performed to evaluate the feasibility of using sodium permanganate to destroy glycolate in the RCT. The results from the laboratory studies were summarized in a series of reports. Reference 9 is a report of lab scale processing of actual DWPF Slurry Mix Evaporate Condensate Tank (SMECT) and Offgas Condensate Tank (OGCT) samples in the SRNL Shielded Cells. Tests at caustic conditions demonstrated sodium permanganate was effective in converting glycolate to oxalate, and permanganate (Mn 7+ ) is reduced to manganate (Mn 6+ ) with no significant formation of carbon dioxide or carbonate. Equation (1) was found to best describe the observed reaction of glycolate with permanganate under nominal (60 to 145 mg/L in RCT) glycolate entrainment conditions.

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The Ambivalent Role of Participants in Nuclear Waste Decisions - 16360

Today's global nuclear sector has entered a critical juncture. Waste management solutions have the potential to influence the prospects of whether new plants will be built and when aging plants will be decommissioned. Even in countries without nuclear power plants, the siting of nuclear waste can present communication challenges and opportunities for public involvement. As new nuclear programs withstand scrutiny, local communities share their want for solutions to nuclear waste before new plant construction commences. Developing nations have begun to explore the potential of adding nuclear power production to their energy mix. They carefully watch established nuclear players to accelerate their plans with lessons learned elsewhere. Some newcomers to nuclear energy will make decisions on new construction based on other operators' abilities to reconcile public concerns regarding radioactive waste. Distinct forms of government, cultural norms and means of public discourse present innovative approaches and changing expectations worldwide. This paper explores how practitioners are evaluating ways to activate stakeholders on nuclear waste siting in nations embarking on new nuclear programs. It investigates the incentives that people have to participate and legitimize the formation of laws, regulations and standards for nuclear materials management. Finally, it highlights examples of public participation techniques from established operating environments and posits that overly prescriptive requirements for stakeholder engagement can reduce the authenticity of public involvement and weaken confidence among participants in the decision making process. (authors)

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Implementing a Low-Activity Waste Direct Feed Capability in the Waste Treatment and Immobilization Plant Project - 15594

The current design of the Waste Treatment and Immobilization Plant (WTP) requires that all waste be processed through the Pretreatment (PT) Facility. The U.S. Department of Energy (DOE) has developed a strategic, phased approach to completing the WTP Project, beginning with feeding some liquid waste from the tank farms to the Low-Activity Waste (LAW) Facility, bypassing the PT Facility. To accomplish this, the DOE Office of River Protection (ORP) will construct a new capability outside WTP to remove cesium and solids from the liquid portion of the tank waste before transferring that waste to the LAW Facility. This new capability, referred to as the Low-Activity Waste Pretreatment System, is outside WTP and not included in the project's scope. The primary driver for the Direct-Feed LAW (DFLAW) approach is to enable commissioning of the LAW Facility for low-activity waste processing before the PT and High-Level Waste Facilities are complete and operational. As such, the functions intended to be provided to the LAW Facility by the PT Facility and the common functionality with the High-Level Waste Facility must be refined and modified. The DFLAW approach will provide necessary systems and facilities to replace these enabling PT Facility functions until the PT Facility is completed and commissioned. In February 2014, ORP requested the WTP contractor to develop separate contract modification proposals for completing: (1) the LAW Facility, Balance of Facilities, and Analytical Laboratory work scope in the current contract through hot commissioning in a manner to accommodate a phased start of the WTP mission; and (2) initial planning and design for incorporating a permanent capability to accommodate DFLAW in the WTP. (authors)

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Evaluation of Thermolytic Production of Hydrogen from Glycolate and Common Tank Farm Organics in Simulated Waste

The Savannah River National Laboratory (SRNL) has performed testing to investigate the thermolytic production of hydrogen gas from organic species present in the Savannah River Site (SRS) Concentration, Storage, and Transfer Facilities (CSTF). SRNL designed an experimental apparatus to measure the time-dependent concentration of hydrogen in the headspace over a reacting mixture of simulated caustic waste media and organic compounds as well as a methodology to calculate a steady-state hydrogen generation rate (HGR) from this data. Special focus was given to glycolate, which is not currently present in CSTF waste but will be contained in the recycle stream from the Defense Waste Processing Facility (DWPF) under the upcoming alternate reductant flowsheet.

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