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Physical and Flow Properties of Glass-Forming Chemicals (V 2 O 5 , SnO, SnO 2 , Cr 2 O 3 , FeCr 2 O 4 , and ZrSiO 4 ) and Mixtures

For an efficient nuclear waste vitrification process at the Waste Treatment and Immobilization Plant (WTP) on the Hanford Site, proper selection and consistent supply of glass-forming chemicals (GFCs) are crucial. Thorough characterization of the GFCs is required to reduce risks in operation of the vitrification facility. Low-activity waste (LAW) will be blended with GFCs to form slurry feeds and then fed to melters and vitrified. To enhance properties of waste glasses, new chemicals are being introduced to the current GFC mixture. In this study, three new GFCs were evaluated for enhanced LAW glass formulations: chromium oxide (Cr 2 O 3 ), vanadium oxide (V 2 O 5 ), and stannic oxide (SnO 2 ). These three oxide components are included in enhanced waste glass formulations, and GFCs with the appropriate physical and flow properties are needed. As a starting point, single metal oxide GFCs, Cr 2 O 3 , V 2 O 5 , and SnO 2 , were sourced and tested. Then, alternative sources of Sn and Cr (SnO and FeCr 2 O 4 ) were tested along with an alternative zircon source (ZrSiO 4 ). This report documents the work performed to collect physical and flow property data on these new GFCs and melter feed slurries generated using these GFCs and simulated low-activity wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Dissolution of Fluoride Salts in Hanford Tank Waste

The Direct Feed High-Level Waste (DFHLW) strategy seeks to bypass the Hanford Waste Treatment and Immobilization Plant Pretreatment Facility while retaining some processing functions to maximize waste feed loading and minimize high-level waste (HLW) waste volume. The DFHLW flowsheet needs leaching, washing, and solids concentration operations either in new or existing tanks. The effectiveness and efficiency of sludge washing has a substantial impact on DST space, mission duration, and the evaporation and low-activity waste (LAW) treatment operations required by these large wash-water additions. One target species requiring washing is fluoride. The HLW glass composition limits for fluorine drive operations to dissolve fluoride-bearing salts into the LAW fraction and thereby maximize waste loading in HLW glass. The fluoride in many high-level wastes at Hanford is predominantly in the form of fluoride-salt precipitates: villiaumite (NaF), kogarkoite (Na 3 FSO 4 ), and natrophosphate (Na 7 F(PO 4 ) 2 ·19H 2 O). Fluoride produces melter off-gas that creates corrosion risk in the off-gas system piping, while the sulfate and phosphate in the fluoride double salts kogarkoite and natrophosphate can be detrimental to glass waste loading. Fluoride salts are sparingly soluble, with solubilities ranging from approximately 40 to 130 kg per kL of pure water, and the dissolution kinetics of the three fluoride salts are not well known. Unexpected delays in a tank dissolution process could be encountered as a result of the lack of information about dissolution rates. In addition, if the double salts show transient non-stoichiometric dissolution of fluoride versus phosphate or sulfate, unexpectedly high concentrations of one of these other constituents could be produced. Washington River Protection Solutions authorized Pacific Northwest National Laboratory to collect the available data for fluoride salt dissolution rate, provide a scoping estimate of dissolution time if possible, and identify gaps in the understanding and predictive capability for estimating dissolution time. Open literature and Hanford reports were reviewed to document, understand, and (where possible) evaluate limitations on fluoride salt equilibria and dissolution kinetics, including both mass transport and surface reaction rate. Scoping estimates of dissolution time were made for mass-transfer-controlled dissolution of spherical particles of fluoride salts suspended in liquid. This is not the only potential governing mechanism; dissolution could be substantially slower if the surface reaction rate (the rate of release of ions from the surface) is the controlling mechanism. When the minimum amount of water for complete dissolution is used and the slip velocity between the liquid and suspended particles is less than or equal to the terminal settling velocity, the estimated mass-transfer rates allow 0.1 mm particles of fluoride salt to dissolve in minutes at 25 °C in water containing no other dissolved salts. Much larger solids, such as the 6-mm chunks that have been seen in heels, could take a few hours to more than a week to dissolve. The actual dissolution times will depend strongly on the actual slip velocity, the extent of particle suspension, constraint by surface reaction rates, the ratio of solvent to solid, and the presence of common ions that shift the solubility equilibria to restrict dissolution of fluoride salts.

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Development of High Alumina Glass Property Data for Hanford High-Level Waste Glass Models - 16231

Roughly half of the projected Hanford high-level waste batches will have waste loadings limited by relatively high concentration of Al2O3. Individual glasses have been formulated and tested to demonstrate that it is possible to increase the loading of these high-Al2O3 wastes in glass by as much as 50%. To implement such increases in waste loading in the Hanford Tank Waste Treatment and Immobilization Plant, the impact of composition on the properties of high-Al2O3 waste glasses must be quantified in the form of validated glass property-composition models. To collect the data necessary for glass property-composition models, a multi-phase experimental approach was developed. In the first phase of the study, a set of 46 glass compositions were statistically designed to most efficiently backfill existing data in the composition region for high-Al2O3 (15 to 30 wt%) waste glasses. The glasses were fabricated and key glass properties were tested: •Product Consistency Test (PCT) on quench (Q) and canister centerline cooled (CCC) samples •Toxicity Characteristic Leaching Procedure (TCLP) on Q and CCC samples •Crystallinity as a function of temperature (T) at equilibrium and of CCC samples •Viscosity and electrical conductivity as a function of T The measured properties of these glasses were compared to predictions from previously existing models developed over lower Al2O3 concentration ranges. Areas requiring additional testing and modeling were highlighted.

Vienna, John D.↗

Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029

This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

99 GENERAL AND MISCELLANEOUS↗

Effect of Form of Chromium on Spinel Formation

A number of Hanford high level waste (HLW) streams contain sufficiently high concentrations of chromium for it to become the waste loading limiting component in the waste. Chromium has relatively low solubility in borosilicate glasses and leads to the formation of various crystal phases including spinels. The extent of crystal formation must be managed during melter processing to prevent operational issues and potential premature melter failure that can result from excessive accumulation of crystals in the melter. In previous work for the Department of Energy (DOE) Office of River Protection (ORP), high-waste loading glass formulations were developed for representative HLW streams that are high in chromium. Results from those studies suggested that there may be differences in the crystallization behavior between glasses prepared in crucible melts and those prepared in DuraMelter 100 (DM100) joule heated melter runs. Understanding the cause of those differences has the potential to lead to the development of methods to further increase waste loadings for high chromium HLW streams through control of the formation of chromium-bearing crystals. As noted previously, the potential causes include differences in the form of components in the batch, including the form of chromium, as well as redox effects. This report describes the results from work to elucidate the origins of the observed differences and to identify melter feed composition modifications to control the formation of chromium-bearing crystal phases. The work focused on determining the effects of the form of chromium and the concentrations of oxidants in feed batches on spinel crystallization. Also evaluated was the use of potential melter feed additives to influence the redox state of chromium in the melt and its effect on spinel crystallization. As described in the Test Plan for this work, the objectives were accomplished through a combination of crucible-scale tests, small-scale tests with additives to influence redox state, and melter tests. This work builds on previous work performed at the Vitreous State Laboratory (VSL) of The Catholic University of America (CUA) for ORP to increase the waste loadings and processing rates of HLW glass formulations.

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Laboratory-Scale Melter Systems for Simulated and Real AP-107 Tank Waste - 20483

The Hanford Waste Treatment and Immobilization Plant (WTP) will process and stabilize waste that is stored in underground tanks on the Hanford Site. At the WTP, the tank waste will be combined with glass-forming and modifying additives to form an aqueous and solid mixture called melter feed, which will then be vitrified in joule-heated melters. A small-scale melter system, the continuous laboratory-scale melter (CLSM), is designed to operate with a continuous feeding process while periodically pouring glass product and collecting off-gas. Two CLSMs were constructed: one in a non-radioactive environment for processing waste simulants and another in a fume hood capable of handling radioactive material for processing actual Hanford tank waste. The waste sample received for vitrification in the CLSM was from Hanford tank 241-AP-107. Approximately 12 liters of melter-feed slurry was vitrified into 7 kilograms of glass product over the course of 10 hours of charging to the CLSM. The partitioning of technetium-99 in the product streams was used to calculate the technetium-99 recovery and help determine when the vitrification process had reached a steady state. The average single-pass retention of technetium-99 in the glass product was then calculated for the processing of AP-107 waste in the CLSM system. (authors)

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Development of High Alumina Glass Property Data for Hanford High-Level Waste Glass Models - 16231

Roughly half of the projected Hanford high-level waste batches will have waste loadings limited by relatively high concentration of Al{sub 2}O{sub 3}. Individual glasses have been formulated and tested to demonstrate that it is possible to increase the loading of these high-Al{sub 2}O{sub 3} wastes in glass by as much as 50%. To implement such increases in waste loading in the Hanford Tank Waste Treatment and Immobilization Plant, the impact of composition on the properties of high-Al{sub 2}O{sub 3} waste glasses must be quantified in the form of validated glass property-composition models. To collect the data necessary for glass property-composition models, a multi-phase experimental approach was developed. In the first phase of the study, a set of 46 glass compositions were statistically designed to most efficiently backfill existing data in the composition region for high-Al{sub 2}O{sub 3} (15 to 30 wt%) waste glasses. The glasses were fabricated and key glass properties were tested: - Product Consistency Test (PCT) on quench (Q) and canister centerline cooled (CCC) samples - Toxicity Characteristic Leaching Procedure (TCLP) on Q and CCC samples - Crystallinity as a function of temperature (T) at equilibrium and of CCC samples - Viscosity and electrical conductivity as a function of T The measured properties of these glasses were compared to predictions from previously existing models developed over lower Al{sub 2}O{sub 3} concentration ranges. Areas requiring additional testing and modeling were highlighted. (authors)

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Low and intermediate level radioactive waste processing in plasma reactor

Methods of low and intermediate level radioactive waste processing comprise: cementation, bituminization, curing in polymer matrices, combustion and pyrolysis. All these methods are limited in their application in the field of chemical, morphological, and aggregate composition of material to be processed. The thermal plasma method is one of the universal methods of RAW processing. The use of electric-arc plasma with mean temperatures 2000 - 8000 K can effectively carry out the destruction of organic compounds into atoms and ions with very high speeds and high degree of conversion. Destruction of complex substances without oxygen leads to a decrease of the volume of exhaust gases and dimension of gas cleaning system. This paper presents the plasma reactor for thermal processing of low and intermediate level radioactive waste of mixed morphology. The equipment realizes plasma-pyrolytic conversion of wastes and results in a conditioned product in a single stage. As a result, the volume of conditioned waste is significantly reduced (more than 10 times). Waste is converted into an environmentally friendly form that suits long-term storage. The leaching rate of macro-components from the vitrified compound is less than 1.10{sup -7} g/(cm{sup 2}.day). (authors)

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Near-Net-Shape Processing of Sintered Fibrous Ceramics Achieved

A variety of sintered fibrous ceramic (SFC) materials have been developed over the last 50 years as thermal barrier materials for reentry applications. SFC materials typically exhibit very low thermal conductivities combined with low densities and good thermal stability up to 2500 F. These materials have flown successfully on the space shuttle orbiters since the 1960's. More recently, the McDonnell Douglas Corporation successfully used SFC tiles as a heat shield on the underside of its DC X test vehicle. For both of these applications, tiles are machined from blocks of a specific type of SFC called an alumina-enhanced thermal barrier (AETB). The sizes of these blocks have been limited by the manufacturing process. In addition, as much as 80 to 90 percent of the material can be lost during the machining of tiles with significant amounts of curvature. To address these problems, the NASA Glenn Research Center at Lewis Field entered a cooperative contract with the Boeing Company to develop a vacuum-assisted forming process that can produce large (approximately 4 square feet), severely contoured panels of AETB while saving costs in comparison to the conventional cast-and-machine billet process. For shuttle use, AETB is slurry cast, drained, and fired to form square billets conforming to the shape of the filtration box. The billets are then cut into tiles of the appropriate size for thermally protecting the space shuttle. Processing techniques have limited the maximum size of AETB billets to 21.5 square inches by 6.5-in. thick, but the space shuttles use discrete heat shield tiles no more than 8 to 12 square inches. However, in other applications, large, complex shapes are needed, and the tiling approach is undesirable. For such applications, vacuum-assisted forming can produce large parts with complex shapes while reducing machining waste and eliminating cemented joints between bonded billets. Because it allows contoured shapes to be formed, material utilization is inherently high. Initial estimates show that the amount of material lost during machining can be reduced by 50 percent or more. In addition, a fiber alignment favorable for minimum heat transfer is maintained for all panel shapes since the fibers are aligned parallel to the contoured surface of the forming tool or mold. The vacuum-assisted forming process can complete the entire forming operation in a matter of minutes and can produce multiple parts whose size is limited only by the size of the forming tool. To date, panels as large as 2 square feet have been demonstrated The vacuum-assisted forming process starts with the fabrication of a permeable forming tool, or mold, with the proper part contour. This reusable tool is mounted over an internal rib support structure, as depicted in the diagram, such that a vacuum can be pulled on the bottom portion of the tool. AETB slurry is then poured over and around the tool, liquid is drawn from the slurry, and the part forms over the tool surface. The part is then dried, fired, and finished machined. Future plans include an evaluation of the need for additional coatings and surface-toughness treatments to extend the durability and performance of this material.

Angel, Paul W.↗

Vitrification of Inactive Incineration Ash in a Full Industrial Scale In Can Melter - 20502

In the framework of the H2020 THERAMIN project funded by the European Commission, CEA performed a pilot-scale trial consisting of the vitrification of inactive ash (Al, Ca, Si, Zn, Bi, K, Mg, etc.) using the In- Can Melter technology. These ashes were coming from incineration of simulated technological wastes such as cotton, latex, neoprene, polyethylene. The borosilicate glass feasibility was first evaluated at a laboratory scale (10-100 g of glass). The volatile behavior of fly ash was controlled by pelletizing to avoid pipes clogging when performing the pilot scale test. A 10 wt.% inorganic binder was added to obtain cohesive pellets. The glass frit composition was selected and its ratio was fixed at 50 wt.%. After this step, a bench test (1 kg of glass) was done to verify the absence of container corrosion. Finally, it was followed by a pilot scale test (50 kg of glass) with 75,000 ash pellets used. The same glass microstructure and chemical composition were observed at each scale. Typical crystals of apatites (Ca, P), zincochromites (Zn, Cr, Al) and bismuth alloys (Bi, Sb) dispersed into a borosilicate glass matrix were observed. The chemical durability of the wasteform was measured during 28 days in a 90 deg. C water. Although there were durable crystalline phases, the hydrolysis rate of the vitreous part is relatively high because of the B and Na contents. However, leach rate profiles suggest that an alteration layer will form, leading to an alteration rate decrease. (authors)

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

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

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Microstructural Characteristics of U-Zr Ingot Slug and Atomized Powder for SFR metal Fuel

Rod-type fuel slugs and spherical powder of U- 10wt.%Zr alloy was also fabricated and characterized in microstructure by injection casting, gravity casting and centrifugal atomization process. U grains became finer from ∼30μm in conventionally ingot rod to ∼2μm in atomized powder. Laminar structure became so fine from ∼0.2μm in conventionally cast ingot rod to 0.05∼0.1μm in atomized powder. Metallic fuels such as U-Zr and U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR), related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's. Metallic fuel slugs of U-Pu-Zr alloys for SFR have conventionally been fabricated by an injection casting method. U-Pu-Zr alloy melts are injected and cast into quartz tube molds during injection casting. Metallic fuel slugs are collected by fracturing quartz tubes from the molds. Demolded quartz tubes are treated with radioactive wastes. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. The metallic fuel for SFR should be remotely fabricated under a radiation shielded environment such as a hot cell, because it contains highly radioactive MA such as Np, Am and Cm. The addition of MA in the fuel composition introduces new challenges in the manufacturing of a metallic fuel slug. Am has a high vapor pressure under a vacuum atmosphere at above a melting temperature of U-Zr or U-Pu-Zr. Used graphite crucibles and quartz molds cannot be recycled easily because of difficult decontamination of radioactive residues on their surfaces. It is necessary to minimize the vaporization of Am and volume of radioactive wastes. Atomized metallic fuel does not need these quartz molds during fabrication, as uranium alloys are prepared as a particle form. It is possible to reduce the formation of radioactive wastes by applying the fabrication process of a particulate metallic fuel. Metallic fuel having a fine microstructure, which originates from a rapid solidification effect during atomization process, results in a higher FG release rate during irradiation. In the case of applying a new forming method such as vibrocompaction in Zr sheath, the sheath can act as a reliable FCCI (fuel/cladding chemical interaction) barrier and a re-constraint for external elongation. In addition, atomized fuel powder, dispersed in a Na liquid metal having a high thermal conductivity, may have a lower temperature distribution during irradiation. In order to develop an efficient fabrication process of metallic fuel for SFR, U-Zr ingot rods were fabricated in a lab-scale with various casting methods. Rod-type U-Zr fuel slugs were fabricated by gravity casting and injection casting. Spherical powder of U-Zr alloy was also fabricated by centrifugal atomization. And then, the microstructural characteristics were investigated by using electron microscopes and energy-dispersive electron X-ray spectroscopy. Atomized U-10Zr powder generally showed spherical morphology of below 200 μm in particle size. U grains became finer from ∼30 μm in conventionally ingot rod to ∼2 μm in atomized powder. Laminar structure became so fine from ∼0.2 μm in conventionally cast ingot rod to 0.05∼0.1 μm in atomized powder. Atomized particulate metallic fuel for SFR has a fine microstructure, resulting in a higher fission gas release rate during irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Mineralization of alkaline waste for CCUS

Ex-situ mineralization processes leverage the reaction of alkaline materials with CO 2 to form solid carbonate minerals for carbon capture, utilization, and storage. Annually, enough alkaline waste is generated to reduce global CO 2 emissions by a significant percentage via mineralization. However, while the reaction is thermodynamically favorable and occurs spontaneously, it is kinetically limited. Thus, a number of techniques have emerged to increase the efficiency of mineralization to achieve a scalable process. In this review, we discuss mineralization of waste streams with significant potential to scale to high levels of CO 2 sequestration. Focus is placed on the effect of operating parameters on carbonation kinetics and efficiency, methods, cost, and current scale of technologies.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Recycled tetrahedron-like CuCl from waste Cu scraps for lithium ion battery anode

Highlights: • High EVA CuCl was recycled with 88.7% recovery from waste Cu craps. • The recycled CuCl displayed regular tetrahedron-like morphology. • The evolution mechanism of the recycled CuCl crystal was proposed. • The corresponding discharge capacity in LIB was 171.8 mAh/g after 50 cycles. • The results shed a new light on resource recovery and environmental protection. - Abstract: The wide applications of metal Cu inevitably resulted in a large quantity of waste Cu materials. In order to recover the useful Cu under the mild conditions and reduce the environmental emission, waste Cu scraps were recycled in the form of CuCl powders with high economic value added (EVA) via the facile hydrothermal route. The recycled CuCl powders were characterized in terms of scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDX) and X-ray diffraction (XRD). The results suggested that the recycled CuCl powders consisted of many regular tetrahedron-like micro-particles. Furthermore, in order to reduce the cost of lithium ion battery (LIB) anode and build the connection of waste Cu scraps and LIB, the recycled CuCl powders were evaluated as the anode active material of LIB. As expected, the reversible discharge capacity was about 171.8 mAh/g at 2.0 C even after 50 cycles, implying the satisfactory cycle stability. Clearly, the satisfactory results may open a new avenue to develop the circular economy and the sustainable energy industry, which would be very important in terms of both the resource recovery and the environmental protection.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Hot-Isostatic Pressing of Sellafield SIXEP Sand/Clinoptilolite Wastes - 18281

Clinoptilolite is an ion-exchange material currently in use at the SIXEP plant (Sellafield, UK) to remove radioactive Cs and Sr from aqueous solution. Clinoptilolite material exists in a granular form that requires consolidation prior to disposal in a Geological Disposal Facility (GDF). The existing baseline treatment technology is to produce cementitious waste packages, which is not ideal due to the large increase in waste volume and poor quality of the wasteform produced. Hot-Isostatic Pressing (HIPing) has been investigated as an alternative treatment method to immobilise this material. This paper presents the results of an investigation that illustrates the capability to process clinoptilolite into a durable, monolithic glass-ceramic wasteform, which should be considered suitable for GDF disposal. The product of HIPing SIXEP sand /clinoptilolite materials is a dense vitreous wasteform with a small quantity of crystalline inclusions, primarily in the form of quartz/cristobalite. Under PCT testing the wasteform produced was shown to be at least an order of magnitude more durable than glasses currently used for the immobilisation of High-Level Wastes (HLW). It was calculated that the treatment of SIXEP sand/clinoptilolite wastes using HIP technology will provide at least a 6.3-fold reduction in the number of waste packages which require management and disposal. (authors)

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Functional magnetic microspheres

Functional magnetic particles are formed by dissolving a mucopolysaccharide such as chitosan in acidified aqueous solution containing a mixture of ferrous chloride and ferric chloride. As the pH of the solution is raised magnetite is formed in situ in the solution by raising the pH. The dissolved chitosan is a polyelectrolyte and forms micelles surrounding the granules at pH of 8-9. The chitosan precipitates on the granules to form microspheres containing the magnetic granules. On addition of the microspheres to waste aqueous streams containing dissolved ions, the hydroxyl and amine functionality of the chitosan forms chelates binding heavy metal cations such as lead, copper, and mercury and the chelates in turn bind anions such as nitrate, fluoride, phosphate and borate.

Yen, Shiao-Ping S.↗

Preferred Concepts for Disposal of the UK Inventory of Depleted, Natural and Low Enriched Uranium - 16382

Depleted, natural and low enriched uranium (DNLEU) forms a significant quantity of the UK inventory of higher activity radioactive materials. The great majority comprises uranic materials produced from uranium enrichment operations and from reprocessing of spent nuclear fuels. The most recent estimate of the total quantity (stocks plus arisings) of UK-owned DNLEU from civil nuclear operations is 170,000 tU. Current UK Government strategy is for DNLEU to be stored indefinitely in existing or new facilities, as required. However, Government strategy for long-term management of uranic materials is under review and a change could potentially cause this material to be reclassified as a waste. An Integrated Project Team (IPT) was formed in 2012 with the objective of identifying and addressing the issues associated with the potential need to dispose of DNLEU, in order to inform wider decision-making on the long-term management of UK DNLEU. Key technical issues considered by the IPT relate to the following: 1. The need for an improved understanding of the UK DNLEU inventory. 2. The identification of preferred chemical and physical form(s), and preferred packaging and disposal concepts for DNLEU - the main subject of this paper. 3. The approach to safety assessment of DNLEU disposal. This paper summarises the disposal concepts under consideration by the IPT, and the rationale for identification of preferred concepts for disposal of DNLEU inventory components. The conclusions will be of interest to other programmes and countries with a requirement to manage large quantities of similar materials. (authors)

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Removal of 14 C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

The aim of the research presented here was to identify the chemical form of 14 C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14 C, with a half-life of 5730 years.

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