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The National Nuclear Laboratory's Approach to Processing Mixed Wastes and Residues - 13080

The National Nuclear Laboratory (NNL) treats a wide variety of materials produced as by-products of the nuclear fuel cycle, mostly from uranium purification and fuel manufacture but also including materials from uranium enrichment and from the decommissioning of obsolete plants. In the context of this paper, treatment is defined as recovery of uranium or other activity from residues, the recycle of uranium to the fuel cycle or preparation for long term storage and the final disposal or discharge to the environment of the remainder of the material. NNL's systematic but flexible approach to residue assessment and treatment is described in this paper. The approach typically comprises up to five main phases. The benefits of a systematic approach to waste and residue assessments and processing are described in this paper with examples used to illustrate each phase of work. Benefits include early identification of processing routes or processing issues and the avoidance of investment in inappropriate and costly plant or processes. (authors)

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U.S. department of energy safety design strategy review and approval beyond critical decision 3 for the high-level waste facility - 15525

The Waste Treatment and Immobilization Plant (WTP) in Hanford, Washington, is the nation's largest environmental remediation construction project. In 2012, at the WTP's High-Level Waste (HLW) Facility, the U.S. Department of Energy (DOE), Office of River Protection (ORP) restricted construction work at the Hazard Category 2 nuclear facility because of unresolved technical issues associated with the facility and of misalignment of the design and nuclear safety basis. By 2013, as additional issues were discovered, ORP limited HLW design and procurement activities. Addressing the technical issues and the misalignment of the design basis with the Preliminary Documented Safety Analysis (PDSA) is a priority for ORP. A proposal was developed to use the concept of a safety design strategy (SDS), as documented in DOE-STD-1189-2008, Integration of Safety into the Design Process, as a planning mechanism to support facility design and PDSA alignment. The standard provides the processes to develop an SDS that will ensure safety is integrated into design early in the project, by the start of preliminary design. Ultimately, as design progresses, an SDS is intended to support development of the project's PDSA. However, in this unique case, the HLW Facility design and construction phase has progressed beyond Critical Decision 3 and DOE-STD-1189-2008 was issued in 2008; 6 years after ORP approved the first WTP PDSA. In July 2013, ORP and the contractor recognized that the unresolved technical issues and design - safety basis misalignment should be addressed through a renewed look at nuclear safety requirements, and agreed that developing an SDS tailored to the criteria in DOE-STD-1189-2008 provided the best path forward to better integrate design and safety. Development of the SDS for the HLW Facility was one of many goals ORP established in a plan that describes how ORP will determine whether, and under what conditions, the contractor should be authorized to proceed with production engineering, procurement, and construction activities for the HLW Facility. Over the next several months, the contractor developed the draft SDS, identifying design basis accidents and control strategies for those accidents. ORP monitored progress and, when appropriate, provided recommendations. In April 2014, the contractor provided a draft SDS to ORP for initial review by an independent DOE Review Team. The Review Team was comprised of representatives from nuclear safety, criticality safety, engineering, fire protection, the Office of Environmental Management, the Assistant Manager WTP, and the DOE Chief of Nuclear Safety. A unique draft document review and interactive review process, facilitated through work sessions and continuous communication with the contractor, culminated in the development of 147 technical comments on topics varying from SDS implementation to hydrogen deflagration in vessels. Comment resolution sought to improve project clarity, delineation of control strategies, and, holistically, facilitate integration of safety into design in the SDS, as it will guide the restart of the HLW Facility production engineering and limited procurement and construction activities. All comments were resolved in support of a contractor submission of the final SDS to ORP at the end of June 2014. Successful resolution of these comments was achieved by continuing the real-time response and incorporation interaction with the contractor to ensure that mutual objectives were satisfied. The response times were facilitated through telephone, E-mail, and a shared file server accessible to contractor and DOE reviewers across the DOE complex. As a measure of success of this process, the Safety Basis Approval Authority (ORP Manager) and the WTP Federal Project Director approved the SDS on August 1, 2014, with no conditions of approval. DOE and the contractor recognized that it would be difficult to implement the SDS without changes to the existing contractor's engineering and nuclear safety processes. Accordingly, the contractor developed or revised a significant number of engineering and nuclear safety plans and procedures to implement the SDS into the design process. Over the coming year, the contractor will conduct a gap analysis between the SDS and the presently approved PDSA to incorporate the design basis accidents and preferred control strategies developed from the SDS into the PDSA. In addition, ORP will conduct an assessment of the contractor's implementation of the SDS into the design process at the appropriate stage of implementation, within 6 to 12 months of approval. Although the SDS was developed late in the HLW project timeline, the document will facilitate alignment of design and the safety basis, and is a key step in the return to full production engineering for the HLW Facility. (authors)

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FX Hg Fogging Fixative Deployment for Mercury Vapor Suppression - 20410

Idaho National Laboratory's (INL) FX Hg fixative solution was deployed at the Y-12 National Security Complex in Oak Ridge, Tennessee to support disposal of mercury-contaminated metal debris. The fixative was dispensed via fogging. Fogging infiltrates non-line-of-sight areas improving fixative coating on complex geometries such as debris piles. FX Hg is the mercury vapor controlling derivative of INL's FX2 fixative. FX2 was jointly developed with the National Nuclear Laboratory of the United Kingdom, and INL developed the FX Hg derivative to suppress mercury vapor generation. The Y-12 deployment was performed in concert with UCOR (URS CH2M Oak Ridge), the cleanup contractor for Y-12. A dumpster filled with debris was fogged with FX Hg. The debris was acceptable as municipal landfill waste, save for the mercury vapor levels measured. If the mercury vapor generation rate could be sufficiently reduced, disposal costs for this waste would be dramatically reduced. FX Hg had previously proven effective at significantly reducing mercury vapor generation rates in bench scale testing at INL. This deployment was the first field-scale deployment of the method. Efficacy results were underwhelming, but interpretation of the outcome is complicated by a paucity of hard data. Methods for improving data capture and analysis are analyzed. Operational difficulties associated with scaling up the process and process improvements for future deployments are discussed. (authors)

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Evaluation of a Sulfate Solubility Limit Greater Than 0.65 Weight Percent in Sludge Batch 10 Glasses

Previous laboratory-scale crucible testing with batch chemicals confirmed that the sulfate (SO 4 2- ) limit for Sludge Batch 10 (SB10) was 0.65 weight percent (wt.%) in glass. This limit signifies that 0.65 wt.% SO 4 2- can be retained in the glass without the formation of a sulfate phase. The Defense Waste Processing Facility imposes this constraint in the Material Tracking Program. Based on preliminary calculations to support the Material Tracking Program, it was anticipated that transfer volumes of the monosodium titanate/sludge solids (MST/SS) stream from the Salt Waste Processing Facility (SWPF) may need to be reduced to maintain projected sulfate concentrations below 0.65 wt.% in glass. Savannah River Mission Completion requested that the Savannah River National Laboratory perform additional sulfate testing to determine whether a sulfate solubility limit greater than 0.65 wt.% is feasible for SB10, which could allow for higher transfer volumes of the MST/SS stream. This report documents the results of the testing at higher sulfate concentrations for the glass composition region defined by the most recent SB10 projection (November 2022) and Frits 473 and 625. Frit 473 was recommended for SB10 and Frit 625 was used during SB9 processing and the SB9 to SB10 transition. A total of twenty-one glass compositions were developed based on the expected compositional variables, which include sludge-only (SO) and coupled processing with the SWPF, waste loading (WL), and frit composition. The target sulfate concentrations were varied from 0.65-0.85 wt.% at 32 and 40% WL. Each glass was prepared from reagent grade chemicals and melted at 1150 °C. Visual observations were used to confirm the presence of a sulfate salt phase on the cooled glass surfaces. Representative samples of each glass were submitted for chemical composition analysis by inductively coupled plasma-optical emission spectroscopy and Cs analysis by inductively coupled plasma-mass spectrometry. Overall the majority of mean measured values are consistent with the target values for each major oxide of interest with less than 5% error. The percent errors for the measured SO 4 2- concentrations are generally less than 10%, which is comparable with previous sulfate solubility study measurements and acceptable. Only the SO glasses based on Frit 625 formed a sulfate phase at a 0.80 wt.% SO 4 2- target concentration at both 32 and 40% WL. The remainder of the glasses did not form a sulfate layer. Due to the formation of the sulfate phase, the limit is conservatively set at 0.70 wt.% based on the measured sulfate concentrations of 0.71 wt.% and 0.75 wt.% for these two glasses. None of the SO or coupled operation glasses based on Frit 473 formed a sulfate salt phase, which supports a sulfate limit of 0.80 wt.%. The following SO 4 2- concentration limits are recommended during SO and coupled SB10 processing: (1) 0.70 wt.% during processing with Frit 625, and (2) 0.80 wt.% during processing with Frit 473.

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CEA decommissioning and clean-up activities legacy waste management and lessons learned - 15506

The CEA has generated radioactive waste in both liquid and solid form for more than 50 years within the scope of its R and D, clean-up and dismantling programs. The vast majority of the waste produced is transferred to disposal centers in approved waste packages. Nonetheless, a large quantity of legacy waste still remains in ageing facilities in which the storage conditions are sometimes no longer satisfactory with respect to the safety requirements that are constantly evolving. As part of a broad clean-up and dismantling process focusing on these ageing facilities managed by the CEA, the top priority is to shut down nuclear activities on two symbolic sites: Grenoble and Fontenay-aux-Roses, which are both located in a dense urban environment. After about ten years of dismantling, the nuclear facilities on the Grenoble site are on their way to be completely decommissioned in the regulatory sense of the term. Furthermore, the radwaste issue on this site has raised no specific problems, though it did require developing innovative and sometimes costly waste outlets. The older Fontenay-aux-Roses site is where the first experimental research on minor actinide separation took place. The facilities on this site have reached a stage in dismantling which necessitates the recovery of irradiating legacy waste containing alpha, beta and gamma emitters. The transfer of this waste to interim storage centers more than 600 kilometers away is a decisive factor in reducing the site's nuclear source term and in ensuring the satisfactory progress of the dismantling operations. This paper details the issues raised by the recovery of legacy waste, which requires greater characterization to better understand its radiological and chemical status. This waste also requires better characterization to meet the storage requirements of new sites designed for storage periods exceeding 50 years prior to its final disposal in a deep geological repository. The issue of transporting waste on public roads is specifically covered, along with the strategy defined to securely transfer all legacy waste resulting from the CEA's very diverse R and D programs. Emphasis is placed on finding the right balance between collecting the necessary data on the waste to be transported, ensuring the reconditioning of this waste so it can be transported safely, and maintaining an efficient industrial-scale waste transfer process to limit the lead times and costs related to legacy waste recovery operations. (authors)

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Evaluation of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

Per a directive from the Department of Energy Savannah River Operations Office (DOE-SR) in 2008, the fissile mass loading concentration must remain below 897 g/m 3 in each high-level waste (HLW) glass canister produced by the Defense Waste Processing Facility (DWPF). To support Sludge Batch 5 (SB5) processing, the Savannah River National Laboratory (SRNL) developed a technical basis that facilitates the evaluation of fissile mass loading of the glass product. The calculation is based on the iron (Fe) concentration in the glass as determined by measurements from the Slurry Mix Evaporator acceptability analysis as well as the glass density. In April 2022, a subsequent DOE-SR directive increased the fissile mass loading limit to 2500 g/m 3 beginning with Sludge Batch 11. Thus, the 897 g/m 3 limit still applies to Sludge Batch 10 (SB10) processing.

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Evaluation of a Zirconium Recycle Scrubber System

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from a synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.

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The Greening of a Plutonium Facility through Personnel Safety, Operational Efficiency, and Infrastructure Improvements - 12108

Chemical and metallurgical operations involving plutonium and other hazardous materials account for most activities performed at the Los Alamos National Laboratory's Plutonium Facility (TA-55). Engineered barriers provide the most effective protection from hazardous materials. These safety features serve to protect workers and provide defense in depth against the hazards associated with operations. Although not designed to specifically meet environmental requirements the safety-based design does meet or exceed the requirements of the environmental regulations enacted during and since its construction. TA-55's Waste Services Group supports this safety methodology by ensuring safe, efficient and compliant management of all radioactive and hazardous wastes generated at the TA-55. A key function of this group is the implementation of measures that lower the overall risk of radiological and hazardous material operations. Processes and procedures that reduce waste generation compared to current, prevalent processes or procedures used for the same purpose are identified. Some of these 'Best Practices' include implementation of a chemical control system, elimination of aerosol cans, reduction in hazardous waste, implementation of zero liquid discharge, and the re-cyclization of nitric acid. P2/WMin opportunities have been implemented in the areas of personnel and facility attributes, environmental compliance, energy conservation, and green focused infrastructure expansion with the overall objective of minimizing raw material and energy consumption and waste generation. This increases technical knowledge and augments operational safety. (authors)

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Batch contact testing with variable Na AN-107 and AP-105

The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. In support of the Direct Feed Low-Activity Waste (DFLAW) program for expediting Hanford tank waste supernate treatment, laboratory-scale ion exchange processing using prototypic unit operations was conducted on AW-105 tank waste at the Pacific Northwest National Laboratory Radiochemical Processing Laboratory. This report describes the batch contact testing for tanks AN-107 and AP-105 to help evaluate CST performance on tanks waste supernate after processing it in the ion exchange columns. Batch columns were performed with the wastes at six Na concentrations.

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Literature Review on the Impact of Thermolytic Hydrogen Generation Rate from the Next Generation Solvent

A literature review was undertaken to assess the thermolytic hydrogen generation potential of the Next Generation Solvent (NGS) system, which is planned to replace the Original Caustic-Side Solvent Extraction (CSSX) Solvent at the Salt Waste Processing Facility. NGS uses a new extractant and suppressor and the same modifier and organic carrier solvent as the Original CSSX Solvent at different concentrations.

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April 2025 Semiannual Composite Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from the Salt Waste Processing Facility is sampled semiannually for transfers to the Saltstone Production Facility. Salt solution is treated at the Saltstone Production Facility and disposed of in the Saltstone Disposal Facility. Per request of customer, X-TTR-Z-00027, Revision 0, one Saltstone Disposal Facility waste form was prepared at Savannah River National Laboratory using the Salt Waste Processing Facility Decontaminated Salt Solution and Z-area premix material for the April 2025 semiannual Toxicity Characteristic Leaching Procedure sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request prepared by Savannah River Mission Completion. At 43 days cured, the Saltstone sample was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure. The April 2025 semiannual Cement-Free grout sample met the South Carolina Code of Regulations for Hazardous Waste Management Regulations 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act metals and Underlying Hazardous Constituents, and met the Saltstone Production Facility Waste Acceptance Criteria that was in effect at the time of the sample collection at the Salt Waste Processing Facility.

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Head-end process for the reprocessing of HTGR spent fuel

The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

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Flowsheet for Conversion of Anion Resin from Nitrate to Sulfate Form for Disposal

Nitrate form anion exchange resins are specifically prohibited for disposal in TRU waste. At some DOE sites, spent resin was mixed with cement to immobilize the resin in a monolith with varying success. At SRS starting in the 1980s, spent anion resin has generally been converted to a more chemically stable form or it was digested with permanganate and transferred as a slurry to the HLW tank farm. In the early years of the site it is reported that over 20 metric tons of organic resin were transferred directly to the tank farm without any attempt to digest it. It appears that in the future conversion to a stable sulfate form could well be the favored disposal path for spent nitrate-form anion resin. This report will document the site history of converting between ionic forms (nitrate to sulfate) and will document process flowsheet conditions for performing that conversion.

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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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Development of a Thermodynamic Model for the Hanford Tank Waste Operations Simulator - 12193

The Hanford Tank Waste Operations Simulator (HTWOS) is the current tool used by the Hanford Tank Operations Contractor for system planning and assessment of different operational strategies. Activities such as waste retrievals in the Hanford tank farms and washing and leaching of waste in the Waste Treatment and Immobilization Plant (WTP) are currently modeled in HTWOS. To predict phase compositions during these activities, HTWOS currently uses simple wash and leach factors that were developed many years ago. To improve these predictions, a rigorous thermodynamic framework has been developed based on the multi-component Pitzer ion interaction model for use with several important chemical species in Hanford tank waste. These chemical species are those with the greatest impact on high-level waste glass production in the WTP and whose solubility depends on the processing conditions. Starting with Pitzer parameter coefficients and species chemical potential coefficients collated from open literature sources, reconciliation with published experimental data led to a self-consistent set of coefficients known as the HTWOS Pitzer database. Using Gibbs energy minimization with the Pitzer ion interaction equations in Microsoft Excel,1 a number of successful predictions were made for the solubility of simple mixtures of the chosen species. Currently, this thermodynamic framework is being programmed into HTWOS as the mechanism for determining the solid-liquid phase distributions for the chosen species, replacing their simple wash and leach factors. Starting from a variety of open literature sources, a collection of Pitzer parameters and species chemical potentials, as functions of temperature, was tested for consistency and accuracy by comparison with available experimental thermodynamic data (e.g., osmotic coefficients and solubility). Reconciliation of the initial set of parameter coefficients with the experimental data led to the development of the self-consistent set known as the HTWOS Pitzer database. Using Microsoft Excel to formulate the Gibbs energy minimization method and the multi-component Pitzer ion interaction equations, several predictions of the solubility of solute mixtures at various temperatures were made using the HTWOS Pitzer database coefficients. Examples of these predictions are shown in Figure 3 and Figure 4. A listing of the entire HTWOS Pitzer database can be found in RPP-RPT-50703. Currently, work is underway to install the Pitzer ion interaction model in HTWOS as the mechanism for determining the solid-liquid phase distributions of select waste constituents during tank retrievals and subsequent washing and leaching of the waste. Validation of the Pitzer ion interaction model in HTWOS will be performed with analytical laboratory data of actual tank waste. This change in HTWOS is expected to elicit shifts in mission criteria, such as mission end date and quantity of high-level waste glass produced by WTP, as predicted by HTWOS. These improvements to the speciation calculations in HTWOS, however, will establish a better planning basis and facilitate more effective and efficient future operations of the WTP. (authors)

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New characterisation method of electrical and electronic equipment wastes (WEEE)

Highlights: ► A novel method of characterisation of components contained in WEEE has been developed. ► This technique was applied on several samples generated from different recycling plants. ► Handheld NIR and XRF were used to determine types of plastics and flame retardants. ► WEEE processing flow-sheet was suggested. - Abstract: Innovative separation and beneficiation techniques of various materials encountered in electrical and electronic equipment wastes (WEEE) is a major improvement for its recycling. Mechanical separation-oriented characterisation of WEEE was conducted in an attempt to evaluate the amenability of mechanical separation processes. Properties such as liberation degree of fractions (plastics, metals ferrous and non-ferrous), which are essential for mechanical separation, are analysed by means of a grain counting approach. Two different samples from different recycling industries were characterised in this work. The first sample is a heterogeneous material containing different types of plastics, metals (ferrous and non-ferrous), printed circuit board (PCB), rubber and wood. The second sample contains a mixture of mainly plastics. It is found for the first sample that all aluminium particles are free (100%) in all investigated size fractions. Between 92% and 95% of plastics are present as free particles; however, 67% in average of ferromagnetic particles are liberated. It can be observed that only 42% of ferromagnetic particles are free in the size fraction larger than 20 mm. Particle shapes were also quantified manually particle by particle. The results show that the particle shapes as a result of shredding, turn out to be heterogeneous, thereby complicating mechanical separation processes. In addition, the separability of various materials was ascertained by a sink–float analysis and eddy current separation. The second sample was separated by automatic sensor sorting in four different products: ABS, PC–ABS, PS and rest product. The fractions were characterised by using the methodology described in this paper. The results show that the grade and liberation degree of the plastic products ABS, PC–ABS and PS are close to 100%. Sink–float separation and infrared plastic identification equipment confirms the high plastic quality. On the basis of these findings, a global separation flow sheet is proposed to improve the plastic separation of WEEE.

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Application of Anaerobic Digestion Model No. 1 for simulating anaerobic mesophilic sludge digestion

Highlights: • The behavior of a anaerobic reactor was evaluated through modeling. • Parametric sensitivity analysis was used to select most sensitive of the ADM1. • The results indicate that the ADM1 was able to predict the experimental results. • Organic load rate above of 35 kg/m{sup 3} day affects the performance of the process. - Abstract: Improving anaerobic digestion of sewage sludge by monitoring common indicators such as volatile fatty acids (VFAs), gas composition and pH is a suitable solution for better sludge management. Modeling is an important tool to assess and to predict process performance. The present study focuses on the application of the Anaerobic Digestion Model No. 1 (ADM1) to simulate the dynamic behavior of a reactor fed with sewage sludge under mesophilic conditions. Parametric sensitivity analysis is used to select the most sensitive ADM1 parameters for estimation using a numerical procedure while other parameters are applied without any modification to the original values presented in the ADM1 report. The results indicate that the ADM1 model after parameter estimation was able to predict the experimental results of effluent acetate, propionate, composites and biogas flows and pH with reasonable accuracy. The simulation of the effect of organic shock loading clearly showed that an organic shock loading rate above of 35 kg/m{sup 3} day affects the performance of the reactor. The results demonstrate that simulations can be helpful to support decisions on predicting the anaerobic digestion process of sewage sludge.

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Results of Routine Strip Effluent Hold Tank and Decontaminated Salt Solution Hold Tank Samples from Modular Caustic-Side Solvent Extraction Unit during Macrobatch 5 Operations

Strip Effluent Hold Tank (SEHT) and Decontaminated Salt Solution Hold Tank (DSSHT) samples from several of the "microbatches" of Integrated Salt Disposition Project (ISDP) Salt Batch ("Macrobatch") 5 have been analyzed for 238 Pu, 90 Sr, 137 Cs, and by Inductively Coupled Plasma Emission Spectroscopy (ICPES). The results indicate good decontamination performance within process design expectations. While the data set is sparse, the results of this set and the previous set of results for Macrobatch 4 samples indicate generally consistent operations. The DSSHT samples show continued presence of titanium, likely from leaching of the monosodium titanate in the Actinide Removal process (ARP).

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