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At least 793 records · Page 44

Permanganate Oxidation of Defense Waste Processing Facility (DWPF) Recycle Collection Tank (RCT) Simulants – Protocol Runs for Nominal and Chemical Process Cell (CPC) Foamover Conditions

Savannah River National Laboratory (SRNL) has completed a work scope in support of the Savannah River Remediation (SRR) Nitric-Glycolic flowsheet. Glycolic acid will replace formic acid as a reducing agent in the pre-processing of High-Level Waste sludge performed at the Defense Waste Processing Facility (DWPF). Glycolic acid will perform the same function as formic acid, namely reduction of mercury and adjustment of feed rheology and melter oxidation/reduction potential, while significantly reducing the potential for hydrogen generation in DWPF processing. Development testing has demonstrated glycolic acid virtually eliminates hydrogen production in the pre-processing steps. The Nitric-Glycolic flowsheet significantly improves DWPF’s ability to address one of the Savannah River Site’s key challenges - the incorporation of effluent received from the Salt Waste Processing Facility (SWPF). SWPF will deliver significant effluent volume to DWPF, resulting in a concurrent increase in DWPF effluent returned to the Concentration, Storage and Transfer Facilities (CSTF). This work scope demonstrates that glycolate can be destroyed under the Nitric-Glycolic flowsheet prior to returning the DWPF recycle stream to CSTF. To avoid potential flammability issues due to thermolysis of glycolate in CSTF, SRR tasked SRNL to quantify and mitigate glycolate returns via DWPF’s recycle stream. The development of a strategy for glycolic mitigation was initiated with a system’s engineering workshop. Various chemical and/or physical solutions for how and where to destroy glycolate were considered – consistent with DWPF’s operational capabilities and process requirements. The workshop identified chemical oxidation of glycolate within the DWPF Recycle Collection Tank (RCT) as the most promising option. (The RCT collects offgas condensate from pre-processing, vitrification, and other unit operations performed in DWPF and is the singular return vessel delivering recycle effluent back to CSTF.) A series of down-select experimental tests were performed that showed sodium permanganate was the most effective oxidant. This work was performed to identify an appropriate operational protocol for DWPF implementation. Building upon the downselect experiments, the work targeted an improved description of suitable RCT conditions prior to the strike (pH, etc.), permanganate addition rate, and destruction reaction kinetics. Data collected were also used to determine the best expression of reaction stoichiometry and the effective minimum permanganate addition required.

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Optimization of Saltcake Removal Flowsheet at SRS through Incorporation of Testing and In-Tank Waste Experience

Saltcake removal at SRS may be performed for several reasons: to provide space for evaporator operation (i.e., to precipitate more salt in the drop tank), to provide feed for salt processing (i.e. immobilize the waste), or to remove the salt for tank closure. Many different salt dissolution techniques have been employed in the 40 years that SRS has been performing salt removal, from a basic “Add, Sit, Remove” method (water is added on top of the saltcake and time is allowed for diffusion), to performing interstitial liquid removal, or using mixing devices to promote contact with the liquid. Lessons learned from previous saltcake removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall saltcake removal process. This includes better understanding of salt properties and behavior during dissolution; the primary concerns for salt dissolution are the release of radiolytic hydrogen and criticality prevention (post-dissolution). Recent developments in salt dissolution include the reuse of dilute supernate and a semi-continuous dissolution (SCD) process, where low volume mixing eductors are used to deliver water near the surface of the saltcake at the same rate as the salt solution is removed and transferred to a receipt tank.

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Vadose Zone Model for B-3A/B Pond for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort; carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The parallel version of the Subsurface Transport Over Multiple Phases (STOMP) simulator officially named the exascale Subsurface Transport Over Multiple Phases (eSTOMP), is used to simulate flow and transport for the vadose models. The documentation for the STOMP code is comprehensive. The theoretical and numerical approaches applied in the STOMP code are documented in a published theory guide (PNNL-12030, STOMP Subsurface Transport Over Multiple Phases Version 2.0 Theory Guide). The code has undergone a rigorous verification procedure against analytical solutions, laboratory-scale experiments, and field-scale demonstrations. The application guide (PNNL-11216, STOMP Subsurface Transport Over Multiple Phases Application Guide) provides instructive examples in the application of the code to classical groundwater problems. The user’s guide (PNNL-15782, STOMP: Subsurface Transport Over Multiple Phases Version 4.0: User’s Guide) describes the general use, input file formatting, compilation, and execution of the code. The primary output of the vadose zone modeling is radionuclide transfer rates to the groundwater for input into the P2R model. The rates will be summed over the 100 by 100 m P2R grid cells that fall within the vadose zone model source domain. The Hanford Site Central Plateau was subdivided into 26 individual vadose zone models, with 13 in the 200 East Area and 13 in the 200 West Area. Waste sites that have a completed performance assessment (PA) or past-leak analysis were not included as sources of radionuclides. Instead the vadose zone to groundwater transfer rates of the Environmental Restoration Disposal Facility, Integrated Disposal Facility, US Ecology, and Waste Management Area C (WMA C) PAs and the past-leak analysis for WMA C were used as direct input to the P2R model. Each of the vadose zone models is documented in separate environmental calculation files (ECFs). This ECF describes the B-3A/B Ponds model. The scope of this ECF is to document the development and results of the B-3 A/B Ponds vadose zone model. CP-63515, Model Package Report: Central Plateau Vadose Zone Models, describes the approach, assumptions, process of determining the number of models required and domain of each model, input data, and processing common to all the models.

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Vadose Zone Model for SALDS for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The parallel version of the Subsurface Transport Over Multiple Phases (STOMP1) simulator, officially named the exascale Subsurface Transport Over Multiple Phases (eSTOMP) is used to simulate flow and transport for the vadose models. The documentation for the STOMP code is comprehensive. The theoretical and numerical approaches applied in the STOMP code are documented in a published theory guide (PNNL-12030, STOMP Subsurface Transport Over Multiple Phases Version 2.0 Theory Guide). The code has undergone a rigorous verification procedure against analytical solutions, laboratory-scale experiments, and field-scale demonstrations. The application guide (PNNL-11216, STOMP Subsurface Transport Over Multiple Phases Application Guide) provides instructive examples in the application of the code to classical groundwater problems. The user’s guide (PNNL-15782, STOMP: Subsurface Transport Over Multiple Phases Version 4.0: User’s Guide) describes the general use, input file formatting, compilation, and execution of the code. The primary output of the vadose zone modeling is radionuclide transfer rates to the groundwater for input into the P2R model. The rates will be summed over the 100 by 100 m P2R grid cells that fall within the vadose zone model source domain. The Hanford Site Central Plateau was subdivided into 26 individual vadose zone models, with 13 in the 200 East Area and 13 in the 200 West Area. Waste sites that have a completed performance assessment (PA) or past-leak analysis were not included as sources of radionuclides. Instead the vadose zone to groundwater transfer rates of the Environmental Restoration Disposal Facility, Integrated Disposal Facility, US Ecology, and Waste Management Area C (WMA C) PAs and the past-leak analysis for WMA C were used as direct input to the P2R model. Each of the vadose zone models is documented in separate environmental calculation files (ECFs). This ECF describes the State-Approved Land Disposal Site (SALDS) model. The scope of this ECF is to document the development and results of the SALDS vadose zone model. CP-63515, Model Package Report: Central Plateau Vadose Zone Models, describes the approach, assumptions, process of determining the number of models required and domain of each model, input data, and processing common to all the models.

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K-Basins Sludge Treatment and Packaging at the Hanford Site - 13585

Highly radioactive sludge resulting from the storage of degraded spent nuclear fuel has been consolidated in Engineered Containers (ECs) in the 105-K West Storage Basin located on the Hanford site near the Columbia River in Washington State. CH2M Hill Plateau Remediation Company (CHPRC) is proceeding with a project to retrieve the sludge, place it in Sludge Transport and Storage Containers (STSCs) and store those filled containers within the T Plant Canyon facility on the Hanford Site Central Plateau (Phase 1). Retrieval and transfer of the sludge material will enable removal of the 105-K West Basin and allow remediation of the subsurface contamination plumes under the basin. The U.S. Department of Energy (DOE) plans to treat and dispose of this K Basins sludge (Phase 2) as Remote Handled Transuranic Waste (RH TRU) at the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The K Basin sludge currently contains uranium metal which reacts with water present in the stored slurry, generating hydrogen and other byproducts. The established transportation and disposal requirements require the transformation of the K Basins sludge to a chemically stable, liquid-free, packaged waste form. The Treatment and Packaging Project includes removal of the containerised sludge from T Plant, the treatment of the sludge as required, and packaging of all the sludge into a form that is certifiable for transportation to and disposal at WIPP. Completion of this scope will require construction and operation of a Sludge Treatment and Packaging Facility (STPF), which could be either a completely new facility or a modification of an existing Hanford Site facility. A Technology Evaluation and Alternatives Analysis (TEAA) for the STP Phase 2 was completed in 2011. A Request for Technology Information (RFI) had been issued in October 2009 to solicit candidate technologies for use in Phase 2. The RFI also included a preliminary definition of Phase 2 functions and requirements. Potentially applicable technologies were identified through a commercial procurement process, technical workshops, and review of the numerous previous sludge treatment technology studies. The identified technology approaches were screened using the criteria established in the Decision Plan, and focused bench top feasibility testing was conducted. Engineering evaluations of the costs, schedules, and technical maturity were developed and evaluated. Recommendations were developed based on technical evaluations. The criteria used in the evaluation process were as follows: (1) Safety, (2) Regulatory/stakeholder acceptance, (3) Technical maturity, (4) Operability and maintainability, (5) Life cycle cost and schedule, (6) Potential for beneficial integration with ongoing STP-Phase 1 activities, and (7) Integration with Site-wide RH-TRU processing/packaging, planning, schedule, and approach. The TEAA recommended Warm Water Oxidation (WWO) as the baseline treatment technology and two risk reduction enhancement options for further consideration during development of the process - size reduction and chemical oxidation (Fenton's reagent). The enhancement options would potentially allow a useful reduction in the total operating time required to process the K Basins sludge. The U.S. Department of Energy's Richland Field Office (DOE-RL) has approved this recommended technical approach. The baseline process can be broken down into the following main process steps: (1) STSC transfer from T Plant to the Sludge Treatment and Packaging Facility (STPF). (2) Retrieval of sludge from the STSCs and transfer to the Receipt and Reaction Tank (RRT). (3) Preparation for immobilization by oxidation using heated water (i.e., WWO) for those batches that require it and concentration by evaporating water at about atmospheric pressure in the RRT. (4) Immobilization by using additives to eliminate free liquids and packaging of the treated sludge into drums. (5) Inspection and handling of the filled drums prior to transfer to a separate storage and shipping facility. (6) Handling of vapor, condensate, and other waste streams generated by the process. Each of these steps is discussed in the paper, together with the current state of progress in developing the technology and requirements for continued development. A schematic of the recommended baseline WWO treatment process is given below. (authors)

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Cap and trade schemes on waste management: A case study of the Landfill Allowance Trading Scheme (LATS) in England

Highlights: • LATS has been effective to achieve a reduction of the amount of landfilled waste. • LATS has been one of the few environmental instruments for waste management with a cap and trade methodology. • LATS has achieved to increase recycling of the biodegradable and other waste fractions. - Abstract: The Landfill Allowance Trading Scheme (LATS) is one of the main instruments used in England to enforce the landfill diversion targets established in the Directive 1999/31/EC of the European Parliament and of the Council of 26 April 1999 on the landfill of waste (Landfill Directive). Through the LATS, biodegradable municipal waste (BMW) allowances for landfilling are allocated to each local authority, otherwise known as waste disposal authorities (WDAs). The quantity of landfill allowances received is expected to decrease continuously from 2005/06 to 2019/20 so as to meet the objectives of the Landfill Directive. To achieve their commitments, WDAs can exchange, buy, sell or transfer allowances among each other, or may re-profile their own allocation through banking and/or borrowing. Despite the goals for the first seven years – which included two target years (2005/06 and 2009/10) – being widely achieved (the average allocation of allowances per WDA was 22.9% higher than those finally used), market activity among WDAs was high and prices were not very stable. Results in terms of waste reduction and recycling levels have been satisfactory. The reduction of BMW landfilled (in percentage) was higher during the first seven years of the LATS period (2005/06–2011/12) (around 7% annually) than during the previous period (2001/02–2004/05) (4.2% annually). Since 2008, the significance of the LATS diminished because of an increase in the rate of the UK Landfill Tax. The LATS was suppressed after the 2012/13 target year, before what it was initially scheduled. The purpose of this paper is to describe the particularities of the LATS, analyse its performance as a waste management policy, make a comparison with the Landfill Tax, discuss its main features as regards efficiency, effectiveness and the application of the “polluter pays” principle and finally discuss if the effect of the increase in the Landfill Tax is what made the LATS ultimately unnecessary.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Reactive transport modeling and characterization of concrete materials with fly ash replacement under carbonation attack - 15477

There currently lacks sufficient knowledge to accurately characterize structural, hydraulic, and chemical performance of cementitious materials that are required to contain and shield nuclear waste for thousands of years. There are a number of degradation methods that can affect the integrity of these materials over this time frame including carbonation which lowers the alkalinity of the material and can induce cracking in materials with steel reinforcement. The rate and extent of carbonation in these materials is impacted by the addition of fly ash and its composition which can vary depending on the fly ash source. To characterize the carbonation reaction, the changes in the microstructure and elemental distribution as a function of extent of carbonation and fly ash type were evaluated using scanning electron microscopy (SEM) coupled with energy dispersive x-ray spectroscopy (EDS). Micro-concretes were prepared with no fly ash (control case) and with fly ash of various calcium compositions. The progression of the carbonation front was observed in the backscattered electron images by the formation of calcium carbonate and depth was confirmed with a phenolphthalein pH indicator test. The migration and deposition of chemical species were observed in carbonated regions from EDS data and were explained based on the changes in material alkalinity and solubility determined from EPA Method 1313 (Liquid-solid partitioning (LSP) as a function of eluate pH). The changes in mass transport properties from carbonation were observed using EPA Method 1315 (Mass transfer rates in monolithic and compacted granular materials using semi-dynamic tank leaching procedures). The carbonation reaction in each material was simulated using LeachXS/ORCHESTRA, a reactive transport modeling tool, which provided further evidence of localized deposition and migration of chemical species due to changes in speciation and solubility. (authors)

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Catalytic Enhancement of Inductively Heated Fe 3 O 4 Nanoparticles by Removal of Surface Ligands

Abstract Heat management in catalysis is limited by each material's heat transfer efficiencies, resulting in energy losses despite current thermal engineering strategies. In contrast, induction heating of magnetic nanoparticles (NPs) generates heat at the surface of the catalyst where the reaction occurs, reducing waste heat via dissipation. However, the synthesis of magnetic NPs with optimal heat generation requires interfacial ligands, such as oleic acid, which act as heat sinks. Surface treatments using tetramethylammonium hydroxide (TMAOH) or pyridine are used to remove these ligands before applications in hydrophilic media. In this study, Fe 3 O 4 NPs are surface treated to study the effect of induction heating on the catalytic oxidation of 1‐octanol. Whereas TMAOH was unsuccessful in removing oleic acid, pyridine treatment resulted in a roughly 2.5‐fold increase in heat generation and product yield. Therefore, efficient surfactant removal has profound implications in induction heating catalysis by increasing the heat transfer and available surface sites.

Moura, Natalia S.↗

Operation of Bubblers in the Savannah River Site Defense Waste Processing Facility Melter - 12166

Savannah River Remediation (SRR) LLC acquired the liquid waste contract at the Savannah River Site (SRS) in the summer of 2009. In order to achieve the main goal of the contract, closing of High Level Waste (HLW) tanks, it was necessary to process more waste throughout the SRS liquid waste facilities. The Defense Waste Processing Facility (DWPF) would need to increase its production rate of radioactive waste glass filled canisters as a part of the plan to achieve this commitment. To attain the increased production rate, four bubblers were installed in the DWPF Melter in September 2010 to agitate the DWPF Melter glass pool. The four bubblers were designed to be installed in existing nozzles on the top-head of the DWPF Melter. The design and fabrication of the four (4) bubblers was accomplished through SRR critical subcontractor EnergySolutions LLC. In addition to the existing bubbler design, a new design concept has been approved and is in the process of fabrication. The new design will allow for the lower end (inside melter) of the bubbler to be replaced while the upper end (outside melter) of the bubbler is reused to minimize cost and waste at the DWPF. The bubblers have been operating in the DWPF Melter for approximately 1 year. The originally installed bubbler set was replaced in January 2011. The bubblers were visually examined once removed from the melter and showed minimal signs of wear. Material testing of the Inconel 690 is being performed to determine if the bubblers operational life can be extended. The use of the bubblers has changed the dynamics within the melter glass pool. This is shown through indications that the bubblers have increased the glass pool circulation. Overall, performance of the bubblers has been encouraging and the DWPF Melter has seen a significant improvement in its ability to process waste since the bubbler installation. The installation of the bubblers accomplished the goal of increasing the glass production capability of DWPF Melter. Currently the DWPF Melter is producing glass at a rate of 91 to 93 kg of glass/hour (200 to 205 lbs of glass/hour). As a result of this increased glass production rate the DWPF has been able to produce the most canisters since radioactive operations began in 1996. These accomplishments can be directly attributed to the bubblers ability to increase mixing/convection within the glass melt pool thereby improving the heat transfer from melt pool to the cold cap. Further, this has helped SRR take a significant step forward toward closing the HLW tanks at the aggressive pace set forth in the most recent liquid waste contract at the SRS. As a path forward, the DWPF is continuing work with SRNL to complete the final metallurgical examination of the bubbler samples to help determine methods to increase the design life of the bubblers beyond the current 6-month restriction. Also, the DWPF is continuing work with EnergySolutions on a new bubbler design concept to minimize cost and waste at the DWPF. The new design will allow for the lower end (inside melter) of the bubbler to be replaced while the upper end (outside melter) of the bubbler is reused. This new design is currently in the final design approval phase and EnergySolutions will be selecting a fabricator to begin the work to build bubbler sets 5 and 6 using the new design. (authors)

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US DOE Office of Technology Innovation and Development - Integration of the EM R and D Program in 2012 and Beyond - 12537

Applied research and technology development has the potential to accelerate environmental cleanup and reduce the cost for cleanup and closure of U.S. Department of Energy (DOE) legacy waste sites throughout the United States. Providing the scientific understanding, knowledge, and technologies to enable successful completion of the DOE Office of Environmental Management (EM) mission, the Technology Innovation and Development program is transforming science and innovation into practical solutions for environmental cleanup. Through integration, collaboration, and communication with DOE partner organization, DOE site managers and contractors, these technologies will reduce human health and environmental risk, cost, and time associated with cleanup and closure. The Office of Technology Innovation and Development (OTID) focused efforts in fiscal year 2011 (FY 2011) to a proactive, visionary program balance with integrated, cross-disciplinary applied research and technology development activities. This transition provides the necessary scientific and technical advancements to address near-term needs. In addition, it fills the critical role in providing scientific approaches and advanced technologies that look beyond today's known needs and requirements to provide innovative technologies to make the necessary long-term changes required to facilitate cleanup and bring sites to closure. The outcomes and impacts of this strategy are summarized in the Impact Plan, which describes potential reduction in life-cycle costs through the development and deployment of advanced technologies supporting EM needs associated with waste processing, groundwater and soil remediation, deactivation and decommissioning, and spent nuclear fuel and materials disposition. Additionally, the OTID International Program Strategic Plan 2010-2015 outlines cooperation and collaboration with the international community that has similar nuclear legacy management experience and expertise to foster transfer of best science practices being used in the field. (authors)

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Establishment of Master Plans on Spent Nuclear Fuel in Korea - 17097

This study pertains to the Master Plans for High-Level Radioactive Waste Management prepared by the Korean Government on the basis of the predicted amount of spent nuclear fuel in Korea and the results of public discussion of the issue. The Korean Government has established the Master Plans for High-Level Radioactive Waste Management on the basis of the results of public discussion of the issue to secure the safety of the nationals and resolve the policy uncertainty that has been accumulated for more than 30 years. To provide a legal basis for the implementation of the Master Plans, legislation of the Act on High-Level Radioactive Waste Management Procedures (tentative) will be promoted, and an independent implementation institute, (tentatively named as) the Management Facility Strategy Committee, and an administrative supporting organization, the Planning and Promotion Group, will be organized and operated. This study is expected to transfer experience of and lessons from determining national policy to countries that have not established their final management policy for spent fuel, based on the formation of national consensus. (authors)

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Effects of Polymer Morphology on Solvent and Catalyst Accessibility during Polyethylene and Polystyrene Autoxidation

Efficient catalytic deconstruction of plastics requires facile solvent and catalyst access to polymer substrates to minimize mass transfer effects. Autoxidation using Co(II) acetate, Mn(II) acetate, and a radical carrier in acetic acid is a promising strategy to deconstruct mixed plastic waste, yet the role of polymer morphology in governing solvent and catalyst accessibility remains poorly understood. Here, in situ simultaneous small- and wide-angle X-ray scattering (SAXS/WAXS), complemented by X-ray fluorescence (XRF) imaging and high-pressure differential scanning calorimetry (DSC), were used to elucidate interactions between acetic acid, a Co/Mn catalyst solution, and semicrystalline polyethylene (PE) and amorphous polystyrene (PS) from room temperature to 160 °C. In PE, acetic acid and catalyst access were confined to amorphous regions and cryomilled particle interfaces at room temperature, while crystalline lamellae remained intact after soaking for up to 34 h. Increasing temperature enabled solvent uptake into PE, followed by solvent-assisted softening above 100 °C, and a modest melting-point depression that removed lamellar transport barriers upon melting. Conversely for PS, acetic acid penetrated the glassy polymer without inducing chain mobility until the glass transition was reached, above which the observed structural changes were consistent with enhanced segmental mobility which enabled bulk penetration. These results suggest that polymer morphology and thermally activated physical transitions arising from diffusion and polymer–solvent interactions can influence whether autoxidation of plastics is transport-limited or kinetically controlled, providing a framework for aligning reaction conditions with reaction outcomes.

09 BIOMASS FUELS↗

Heat Transfer from Glass Melt to Cold Cap: Gas Evolution and Foaming

In electric melters, the conversion heat is transferred through the foam layer at the cold cap bottom. Understanding cold cap foaming is thus important for enhancing the efficiency of both commercial and waste glass melters as well as for the development of advanced batch-to-glass conversion models. Observing foam behavior “in-situ” is still impossible, i.e., viewing directly in glass melters. To investigate the feed foaming behavior in laboratory conditions, we employed the feed volume expansion test, evolved gas analysis, and thermogravimetry. Combining these techniques helps assess the cold cap bottom temperature that directly influences the temperature gradient at the melt/cold cap interface, and thus the rate of melting. We also discuss the behavior of cavities formed by coalescing primary foam bubbles and ascending secondary bubbles.

Hrma, Pavel R.↗

Sampling and Analysis Plan (SAP) for Assessment of LANL-Derived Residual Radionuclides in Soils within Revised Tract A-16-a for Land Transfer Decisions

Material Disposal Area (MDA) B is an inactive subsurface disposal site located south of DP Road in Technical Area 21 (TA 21). It is approximately 1600 ft east of the intersection of DP Road and Trinity Drive (Figure 1). Land proposed for for transfer is identified as tract A-16-a and constitutes approximately 30 acres with about 6 of those acres being above the trench from which waste was removed (LANL 2012a).

61 RADIATION PROTECTION AND DOSIMETRY↗

Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy specimens of Midland beltline weld have been machined and just arrived to ORNL as part of this international collaboration. The ORNL will initiate tests of the irradiated Linde 80 weld in FY2017 and results of this international program will be reported in FY2018.

42 ENGINEERING↗

The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

36 MATERIALS SCIENCE↗

USE OF SMALL SPECIMENS FOR FRACTURE TOUGHNESS EVALUATION OF RPV STEELS

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of standard Charpy specimen or its broken half may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs and most likely to be used in advanced reactors as per ASME code.

Sokolov, Mikhail A↗

A gLite FTS based solution for managing user output in CMS

The CMS distributed data analysis workflow assumes that jobs run in a different location from where their results are finally stored. Typically the user output must be transferred across the network from one site to another, possibly on a different continent or over links not necessarily validated for high bandwidth/high reliability transfer. This step is named stage-out and in CMS was originally implemented as a synchronous step of the analysis job execution. However, our experience showed the weakness of this approach both in terms of low total job execution efficiency and failure rates, wasting precious CPU resources. The nature of analysis data makes it inappropriate to use PhEDEx, the core data placement system for CMS. As part of the new generation of CMS Workload Management tools, the Asynchronous Stage-Out system (AsyncStageOut) has been developed to enable third party copy of the user output. The AsyncStageOut component manages glite FTS transfers of data from the temporary store at the site where the job ran to the final location of the data on behalf of that data owner. The tool uses python daemons, built using the WMCore framework, and CouchDB, to manage the queue of work and FTS transfers. CouchDB also provides the platform for a dedicated operations monitoring system. In this paper, we present the motivations of the asynchronous stage-out system. We give an insight into the design and the implementation of key features, describing how it is coupled with the CMS workload management system. Finally, we show the results and the commissioning experience.

Cinquilli, M.↗