Development of Iron Phosphate Waste Forms for Fission Products Waste Streams
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MRWFD Center for Radiation Chemistry Research Activities
Presentation for DOE-NE Aqueous Separation 2021 PIs’ Meeting updating on the FY21 research activities of the INL Center for Radiation Chemistry Research.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Continuous Synthesis of Metal-Organic Frameworks for Xenon Capture
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38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Savannah River Site HB-Line Plutonium Oxide Production Process Simulation Model - 17458
The AFS II/Plutonium Oxide Production process is in its initial operational phase at the Savannah River Site's (SRS) HB-Line. The process is new for the facility where it is being performed, but the facility itself has been in use on and off for over 50 years. Beginning this process inside the facility has resulted in multiple process delays, due to both equipment issues and process restrictions. In order to better understand the process and to identify ways to improve facility performance, Savannah River Nuclear Solutions (SRNS), in conjuncture with Newport News Shipbuilding (NNS), developed the HB-Line Process Simulation Model. This tool completely replicates the operations that comprise the HB-Line process and simulates all of the constraints associated with the process. The model realistically accounts for the number of personnel required for a process step, the time required for a process step to occur, rules related to the parallel operation of process steps, and realistic equipment and operational delay data. The result is a tool that allows SRNS to test different scenarios for improving the process throughput, while realistically accounting for potential process delays and restrictions. The tool is currently in use and has allowed SRNS to present to their customer, the National Nuclear Security Administration (NNSA), estimated production capabilities over wide-ranging timetables and assumptions. (authors)
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Contribution Slides for External IAEA Conference Radiation Chemistry Review
Contribution slides for an invited talk (presented by Professor Mohamad Al-Sheikhly, University of Maryland ) at the International Atomic Energy Agency (IAEA)- Virtual Workshop on Radiation Technology for Industry and Environment. The title of the review talk is "Future Trends in Radiation Technology Applications in Advancing Science and Engineering".
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Developing a new criticality methodology to support decommissioning of legacy plutonium facilities
To decommission a legacy plutonium finishing facility, a number of gloveboxes need to be disposed of. These gloveboxes have been in place since the 1960's and were operational until the 1980's. A period of initial clean out and decommissioning took place in the 1990's and the first gloveboxes were removed from the line in the early 2000's. Compliance with downstream waste handling facility requirements meant that the gloveboxes needed to be removed using intensive 'hands-on' methods in order to size reduce them to fit into 200 litre (55 gallon) drums. This is the standard approach for handling waste items known as Plutonium Contaminated Material (PCM) on the Sellafield site and was credited in the criticality safety case for the decommissioning of the gloveboxes as it allowed each waste drum to be individually assayed such that the fissile content of the waste could be known. Not only was this approach time-consuming and laborious but as it involved operators working with cutting tools in a highly contaminated area it was challenged as being overly restrictive to operations and potentially hazardous to the workers. In order to decommission the remaining nine gloveboxes in the facility, a new decommissioning approach and associated criticality methodology has been developed. Instead of intensive size reduction and waste drum assay, the gloveboxes will be removed with little or no size reduction and put into storage crates without assay. The criticality assessor will review the operational history of each glovebox, along with the decommissioning records and relevant plant data, to make an evidence-based judgement that the crated glovebox represents no greater a residual criticality risk than would have been realised from the equivalent number of PCM drums that the standard decommissioning approach would have generated. This method identifies, in conjunction with the plant personnel, a number of review points to ensure that before the glovebox is removed any bulk residue material has been recovered and no more than a safe amount of moderating material (fixative) has been added to fix any remaining fissile contamination. Finally, once these review points have been confirmed to be complete - through a stakeholder meeting with criticality safety, plant operators and project managers - the glovebox will be placed in a crate, and this crate will be declared a Low Criticality Risk crate. No formal assay will be required prior to exporting the crate to the downstream waste handling facility and no fissile mass value will be assigned to the contents of the crate. No further intrusive work will be required on the crate and it can be safely stored without any spacing requirements, pending final disposal. (author)
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Evaluation of Options for Disposition of Mk-42 Materials - 18034
In the late 1980's, the DOE defense programs identified a need to recover special isotopes of americium, curium, and plutonium from Mark-42 (Mk-42) targets that had been irradiated at the SRS. Eighty Mk-42 target segments were transferred from SRS to ORNL for recovery of these special isotopes. Approximately 65% of the target material has been processed, and the recovered material converted to solids. The remaining Mk-42 materials are stored either as targets in their irradiated forms or as target material from various stages of processing. The Pu, Am-243 and Cm-244 in these targets are valuable materials; for example, the Am/Cm materials are potential feedstocks for production of heavy actinides (Cf-252, Bk-249, etc.). In addition, rare long-lived fission product (LLFP) isotopes present in the targets in much smaller quantities, such as Zr-93, also have value to the research community. Although the isotopes in the Mk-42 materials have value, there is continual pressure to de-inventory the material in storage and only retain the most valuable materials needed to meet mission needs and provide inventory space for other materials to be added as they become available. Evaluations are underway for potential disposition options for the Mk-42 materials. Previous studies have included identification and evaluation of disposition options for the various Mk-42 materials in storage at ORNL [1]. Evaluations have included options for recovering the major isotopes and evaluating the waste that would be sent to disposal facilities. This paper summarizes the next steps in the evaluation process to develop an optimum path forward for the disposition of the Mk-42 materials. It includes an evaluation of the future programmatic need for the materials, the impact of storage on facility operations, and additional evaluations of options for dispositioning the materials. In addition to recovery of the major isotopes previously evaluated, it also looks at the potential to recover rare LLFP isotopes present in the targets in small quantities. (authors)
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An Industrial Perspective of Research Within the DISTINCTIVE Program - 18045
UK Engineering and Physical Sciences Research Council (EPSRC) and three industrial partners, the Nuclear Decommissioning Authority (NDA), the National Nuclear Laboratory (NNL) and Sellafield Ltd (SL) are supporting the DISTINCTIVE (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) consortium of eleven universities to carry out research in four theme areas of importance to the UK nuclear industry: Advanced Gas-cooled Reactor (AGR), Magnox and exotic spent fuels; PuO{sub 2} and fuel residues; legacy ponds and silo wastes; and structural integrity. This paper describes the main industry challenges in each of these theme areas and some of industry led activities to address these challenges, and provides examples of where the DISTINCTIVE research output is assisting the nuclear industry in meeting these challenges. (authors)
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Uranium Ion Exchange Resin Performance - A Case Study at 200 West Pump and Treat - 18103
The 200 West Pump and Treat Facility (200W P and T), located on the U.S. Department of Energy's Hanford site, is a 2,500 gallon per minute (GPM) plant designed to capture and treat an eight square kilometer groundwater plume. The groundwater contamination is largely a result of process liquids disposed from operations involving the separation of nuclear materials and storage of spent nuclear fuel. This document will report on the performance of a full scale uranium ion exchange system used to treat contaminated groundwater and other waste streams. The uranium ion exchange system at 200W P and T was designed as a three vessel operation with a lead, lag and polish vessel, which contain 300 cubic feet of resin in each vessel. The ion exchange system treats a portion of the total plant flow with a maximum capacity of 400 GPM. Prior to start-up, one key issue was the potential to treat feed streams with a wide range of uranium concentrations (22 to 150,000 microgram per liter [μg/L]). The resin change-out frequency was a primary concern. A literature review was performed, resulting in resin capacity and change out frequency that varied by two orders of magnitude (1.2 milligram per gram [mg/g] to 104 mg/g capacity). Since startup in September 2015, the ion exchange system has experienced a dynamic feed stream as wells were brought online, which changed the contaminant concentrations. The plant also treats leachate liquid waste from on-site low-level waste site, known as the Environmental Restoration Disposal Facility (ERDF), which contains uranium as well as other competing anions, such as sulfate. The uranium breakthrough is closely monitored throughout the resin beds with weekly sampling events at the influent, effluent, and intermediate ports in the first vessel. In the influent to the ion exchange system, the uranium concentration fluctuated from 24 μg/L up to 910 μg/L. The information on influent and effluent was used to determine the change out frequency and the capacity of the resin for uranium under these operating conditions. (authors)
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Hot Isostatically Pressed Ceramic Waste forms for High Dose Spent Adsorbents - 18129
Decommissioning and clean-up of nuclear facilities requires the development of new technologies to condition radioactive wastes, producing passively safe waste packages of minimal volume, to reduce storage and disposal costs. We have applied hot isostatic pressing to demonstrate conceptual waste forms for high dose spent adsorbents (i.e. ion exchange materials) present on the Sellafield site, UK, and Fukushima Daiichi site, Japan. These radioactive wastes pose several challenges which demand early conditioning to produce a passively safe wasteform, including: a wet and granular nature, and hence dispersible as a result of loss of containment; a very high dose rates as a result of their selectivity for short lived radionuclides (e.g. Sr-90, Cs-137, Co-60); radiogenic self-heating, as a result of the concentration of such lived radio nuclides; hydrogen production as a result of radiolysis of entrained water [1]. Here, we demonstrate the conversion of natural inorganic ion exchange materials into multiphase ceramic waste forms, achieving a waste loading of 100 wt.% and density in excess of 97% theoretical. High resolution thermogravimetric analysis coupled with mass spectroscopy was utilised to characterise the evolution of water and volatiles during the in-can bake out step, prior to the HIP cycle. This allowed optimisation of the bake out parameters (temperature, time and vacuum) to enable complete removal of water and volatiles, affording ceramic bodies with minimal residual porosity by hot isostatic pressing at 1250 deg. C for 4 h in stainless steel canisters. Characterisation of the ceramic waste forms by SEM /EDX and X-ray diffraction revealed the nuclides of concern to be incorporated within well-known natural mineral and synthetic phases, with Cs partitioning into the K-feldspar. (authors)
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Synthesis and Characterisation of Caesium Phosphomolybdate and Zirconium Molybdate found in the Highly Active Storage Tanks at Sellafield - 18154
Caesium phosphomolybdate (Cs{sub 3}PMo{sub 12}O{sub 40}.xH{sub 2}O, CPM) and zirconium molybdate (ZrMo{sub 2}O{sub 7}(OH){sub 2}.xH{sub 2}O, ZM) are two known fission products that precipitate out within the Highly Active Storage Tanks (HASTs) at Sellafield, potentially causing 'hot-spots' and blocking pipes. Currently, there is not enough behavioural understanding of these compounds present in the Highly Active Liquor (HAL) for effective Post Operational Clean Out (POCO) strategic planning. As the HASTs do not have an in situ monitoring system the behaviour of the HAL cannot be directly observed. Therefore, as with this research, non-radioactive simulants can be utilised to aid behavioural understanding to assist with the planning of future waste management programs. CPM, along with two different morphologies of ZM were characterised and compared to the common oxide particle material titanium dioxide (TiO{sub 2}). The study provides a detailed physical characterisation of the simulants including: particle shape, particle size, water content and crystal structure, along with zeta particle potential, suspension settling and compressive yield stress. CPM was found to be spheroidal in nature with a particle size of 275 nm and an x value of 13 H{sub 2}O which crystallises as a cubic lattice with space group Pn-3m. ZM was found to have a range of morphologies from cuboidal to wheat-sheaf, varying in size from 10 μm to 20 μm, with an x value of between 3 and 4 H{sub 2}O, which crystallise as a body-centred tetragonal lattice with space group I4{sub 1cd}. TiO{sub 2} used as a comparison simulant was found to be spheroidal in shape with particle size of 725 nm. The isoelectric points (IEPs) of the two ZM samples (ZM-A and ZM-B) with different morphologies were found to be pH 2.5, while for TiO{sub 2} it was pH 3.2. For CPM it could not be confidently identified due to its instability within the acidic environment. The settling rates and compressive yield stress of all simulants were tested in both water and 2 M nitric acid, conditions that replicate the HASTs. TiO{sub 2}, ZM-A and ZM-B all found settled faster in acid than in water. For CPM, slightly faster settling in acid was measured. For all samples it was found that as the particle concentration increased the settling rate decreased, which is a result of hindered settling. For compressive yield stress, similar trends were observed for CPM with little variation in its behaviour in acid and water, suggesting that the particles are highly aggregated in both fluids. For ZM samples little change in equilibrium volume fraction was measured with an increase in pressure (both acid and water). However, overall the sediment bed was denser in water, a result of the rapid settling in acid not allowing the non-spherical particles to pack in an ordered manner. It has been highlighted that both CPM and ZM within the acidic conditions of the HASTs, result in unstable and aggregated compounds, which settle quickly to form large open networks. (authors)
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Product Quality Assessment of Glass Formulations for High Sodium Nuclear Waste Streams - 18198
Sodium carbonate is currently being considered as a wash-out reagent for the removal of the settled solids in the unagitated Highly Active Liquor (HAL) storage tanks at Sellafield. As the settled solids are expected to comprise mainly zirconium molybdate (ZM), this will result in a feed to the Waste Vitrification Plant (WVP) containing high concentrations of both molybdenum and sodium. This feed is expected to be challenging for WVP due to solubility limitations in existing base glass compositions and the operating temperature. In particular, depending on the wash-out strategy employed, there is the potential to produce high volumes of sodium-containing waste, which will result in large quantities of vitrified HLW containers. Hence, new glass compositions are being investigated as part of a 3-year Innovate UK-funded collaborative project in order to maximise the sodium content of the vitrified product and minimise the number of containers produced. This paper describes the product quality assessment of several sodium titanosilicate (NaTiSi) compositions that can potentially contain up to ∼15 wt% waste Na{sub 2}O, more than double what is possible using the current UK HLW glass borosilicate composition. (authors)
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Development of Hot-Isostatic Pressing for Treatment of UK ILW Radioactive Wastes - 18280
Hot-Isostatic Pressing (HIPing) has been proposed as an improved method for the treatment of radioactive wastes in the UK. Sludges and slurries housed on the Sellafield site are a prime candidate for HIP treatment due to their large volumes and issues associated with the use of the baseline cementation strategy. Options have been developed by GeoRoc Ltd to co-process these wastes, including a detailed conceptual plant design. This paper will describe research and development aimed at demonstrating the viability of the concept presented in this design. This paper summarizes two studies performed to address the knowledge gaps identified for treatment of these wastes using HIP technology. The first study related to the poorly characterized nature of the wastes, which requires any treatment option to be flexible in terms of the chemical and physical properties of the waste feed. The second study involved demonstrating that HIPing can be effectively utilized at a scale applicable to treat a waste inventory estimated to be in excess of 5,000 m{sup 3}. The first section of this paper presents a summary of research carried out to determine where variation in the waste stream composition may affect the processing and wasteform produced when HIPing Magnox sludges. This study aimed to qualify where restrictions on the process envelope may exist due to poorly defined waste-stream characteristics. A baseline wasteform was selected with the aim of achieving significant waste volume reduction, while producing a physically and chemically stable wasteform. A glass-bonded magnesium-titanate/magnesium-silicate wasteform was selected as the preferred option for the Magnox sludge wastes. A series of studies investigated the densification and phase formation of waste-forms by altering key variables in the Magnox sludge simulant. To test the extremes of the processable waste loading, samples were produced and characterized at 100 wt% waste-loading for Magnox sludge, Site-Ion Exchange Plant (SIXEP) sand/clinoptilolite and a 50:50 mix of the two waste-streams. Further studies involved a) altering the waste loading to additive mass ratio (to simulate changes in the waste loading/sludge solids content), b) varying the particle size distribution from the datum to include only small or large particulates in the sludge feed; and c) introducing up to 20 wt% metallic Magnox alloy, organic species or carbonate materials (to determine how their inclusion may affect the processing and densification of the materials). A U{sub 3}O{sub 8} loaded sample was also produced using the baseline Magnox sludge formulation. All the tests performed were shown to produce monolithic, well densified waste-forms with substantial volume reductions. Minimal variations from the baseline were noted but the waste-forms produced with these deviations was determined to be of comparable quality to of the baseline product. U{sub 3}O{sub 8} was shown to behave in a similar fashion to the ZrO{sub 2} surrogate used to simulate it in the process envelope studies and was encapsulated in the magnesium titanate/magnesium silicate matrix. This work demonstrates that HIP technology can successfully produce waste-forms across the broad process envelope expected in the Magnox sludge inventory. It is also possible to produce monolithic waste-forms at 100 wt% waste loading from Magnox sludges, SIXEP sand clinoptilolite slurries or a co-processed mixture of the two. The second part of this paper summarizes a commercially relevant demonstration of HIP technology for the treatment of sludge and slurry wastes housed at the Sellafield site (UK). This study culminated in production of the largest HIPed wasteform successfully produced to date. This study was performed in a series of stages. Investigations into formulation and properties of waste-forms were performed at a lab scale to evaluate suitable wasteform systems and identify suitable combinations of additives and waste loadings. Scale up demonstrations were then performed in stages with custom-designed HIP canisters. Scale-up and characterization of key wasteform properties were successfully performed with initial canister volumes of 50 ml, 400 ml and 8 L. Suitable front-end unit operations to produce a granular, free-flowing calcined simulant were demonstrated at an engineering scale. The demonstration was successful in processing a HIP canister with an initial starting volume in excess of 100 L. A 75 % volume reduction from the raw sludge was achieved at scale while demonstrating that HIP is applicable at suitably large scale for immobilization of ILW waste. (authors)
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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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Adaptation of Commercially Available Robots for Conditioning and Packaging ILW at Sellafield - 18282
The Legacy Ponds and Silos at Sellafield site contain some of the most hazardous Intermediate Level Waste (ILW) in the UK. As part of a multi-billion pound decommissioning programme, this waste will be retrieved and transported across the site for conditioning and packaging in the Box Encapsulation Plant (BEP). Current plans are to commence plant active commissioning in 2022, which will represent a major milestone in the UK nuclear decommissioning programme. Within the plant, skips of waste will be transferred from the shielded transport package into standard 3 cubic metre ILW containers. Waste from the Pile Fuel Storage Pond and First Generation Magnox Storage Pond will be immobilised in cementitious grout before being transferred to an adjacent store. Waste from the Magnox Swarf Storage Silo will not be immobilised in grout, but will be packaged, providing multi-barrier containment. During the concept design, the project team selected commercially available robots as the most viable technology for carrying out the numerous dexterous and repetitive tasks involved in characterising, conditioning and packaging the waste within the shielded 'hot' cells. Significant technical challenges had to be overcome during the design to adapt a commercially available robot for use within a high radiation / high contamination environment, where human access is not safe even for short periods of time. These challenges included: - Designing robot assemblies that could be retrofitted into the existing cell structure that was constructed as part of an earlier project but never completed; - Designing robot assemblies that could be recovered to their 'transport' position without requiring manual intervention; - Designing robot assemblies that could be remotely removed and replaced with new ones if they failed part way through the plant operational life; - Designing through-wall service and power connections that allow maintenance and recovery; - Designing control systems and human-machine interfaces that enable both manual and automated operation of the robot, whilst avoiding collisions with the in-cell cranes and other in cell furniture; - Assessing the radiation tolerance of all items. Bespoke computer simulations and full-scale development trials have been carried out, using non-active test materials, in order to demonstrate the Technology Readiness Level (TRL) required by Sellafield Limited (SL) at each stage of the project [1]. The results of these trials have been recorded in both video and written form so that they can form part of the project's Knowledge Management database. Detailed designs and specifications for the robot assemblies have been updated to incorporate the recommendations from the development trials. Procurement and manufacture of robot assemblies has already commenced to deliver two waste handling robots, two waste container lidding / swabbing robots and associated critical spares. (authors)
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Miniature Inspection Robot for Restricted Access exploration (MIRRAX) - 18324
The MIRRAX remote inspection vehicle has been developed to explore and characterise enclosed legacy facilities on the Sellafield site which have limited access ports. Its re-configurable design means that it can be deployed through a 150 mm port before changing its footprint shape to provide a stable platform for mobile characterisation. It can support both geometric and radiometric scanning. (authors)
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Robotic Platforms for Remote Access and Sensing Applications for Nuclear Decommissioning - 18325
This paper provides an updated summary of the research being conducted by the University of Manchester Robotics group (UoM Robotics) [1] at the University of Manchester, UK, in the area of remote characterization through the use of robotic systems. The research is in support of the decommissioning plan for the Sellafield nuclear site in west Cumbria, UK and Fukushima Diiachi, Japan. UoM Robotics works in collaboration with a number of companies and research institutes including Sellafield Ltd, the National Nuclear Laboratory, Universities of Lancaster, Birmingham, Bristol Robotics Laboratory, Italian Institute of Technology and the National Maritime Research Institute in Japan. The progress of a number of novel mobile robotic platforms being developed at the University of Manchester is presented in this poster presentation. Also presented is the on-going work on the remote sensing technologies and techniques which will be deployed on the robotic platforms. (authors)
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Cake Permeability of Waste Simulants found in the Highly Active Storage Tanks at Sellafield - 18676
Caesium Phosphomolybdate (CPM) and Zirconium Molybdate (ZM) are two compounds causing issues within the Highly Active Storage Tanks (HASTs) at Sellafield. CPM has the potential to cause 'hot spots' possibly affecting the integrity of the tanks due to its settling and radioactive properties. ZM is thought to cause mobility issues when it is re-suspended by the jet ballasts in the HASTs. CPM and ZM were synthesised following the method published by Paul. The permeability method followed was published by Loginov. Conclusions: CPM has a higher overall cake resistance; Suggests a more open network for the ZM bed than the CPM bed; Cubes would be expected to pack tighter than spheres, suggesting the orientation of the cubes has some effect on the packing; In future the use of 2 M nitric acid instead of water will be explored as this is the natural conditions found in the HASTs.
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