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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Water Cleaning System for the Exhaust Pipe of HLW Glass Melter (LFCM) - 17180

Vitrification technology has been generally applied to stabilize High-Level Radioactive Waste (HLW) generated from spent fuel reprocessing. In Japan, Joule Heated Liquid Fed Ceramic Melter (LFCM) system is selected for HLW vitrification and the recent active test has been successfully demonstrated its stable operation. The higher reliability of off-gas treatment system (OGTS) which consists of wet scrubber, absorbers and filters connected to the LFCM is also important to ensure long term system operation. The OGTS removes radioactive gases, mists and dust generated by LFCM and maintains radioactive containments by keeping negative pressure inside. In the past active and inactive tests, it was observed that differential pressure in the exhaust pipe attached to LFCM gradually increased during continuous operation and some deposits were observed inside the off-gas pipe. So the cleaning operation needs to be improved and more frequent in order to avoid the exhaust pipe clogging. Therefore we investigated several cleaning systems and operational methods for the exhaust pipe of LFCM. As a result, we have successfully developed the modified water cleaning system. In this presentation we will introduce our water cleaning system. In addition, we will present the result of inactive mock-up tests for the visualization of this water cleaning system. (authors)

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DOE OSTI2017

Experience in Decommissioning of Nuclear Fuel Cycle Facilities in Germany - 17195

Several nuclear fuel cycle facilities have been decommissioned in Germany and the sites released from regulatory control. There is an overview of past and current decommissioning projects of fuel cycle facilities, either past fuel element fabrication facilities or past and current (experimental) reprocessing facilities. Several examples of technological advances are given, being developed due to obstacles, which occurred during conduct of these decommissioning projects. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Understanding Uranium Binding Mechanisms and Speciation on Stainless Steel - 17462

Stainless steels have been exposed to nitric acid solutions containing natural uranium under similar conditions to those found in reprocessing plants. Solution and solid analysis of the stainless steel samples has been undertaken to describe the contamination phenomenon. General corrosion of the surface was seen to take place over the month-long experiments. Contamination has been shown on all exposed surfaces. Solution analysis indicates that an equilibrium state is reached after 2 weeks exposure to the 12 M HNO{sub 3}. Surface analysis through autoradiography and X-ray absorption near edge structure (XANES) indicate that multiple mechanisms of uptake may occur and consequently concentrated regions of contamination may be produced. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2017

Pile Fuel Cladding Silo Deflector Plate Removal Project Using Long Reach Tooling and Suspended Abrasive Water Jet Cutting System - 17541

The Pile Fuel Cladding Silo (PFCS) at the Sellafield decommissioning, reprocessing and nuclear waste management facility in the UK was constructed in the late 1940-s and used for the storage of Intermediate Level Waste (ILW) arising from the Windscale Piles, Calderhall and Chapelcross Reactors until the 1960's. The PFCS is 21 metres high and comprises 6 identical tall compartments with the waste added to the silo via a charge-hole at the top of each compartment; each compartment is split by a central dividing wall that runs from the base of the silo to a few metres below the underside of the silo roof. A deflector plate assembly (DPA) supported on two horizontal beams is located between the top of the dividing wall and the roof and allowed the waste to spread as it fell into the silo and prevent excessive mounding. Sellafield Ltd is currently engaged in an overall programme of works to remove the waste for storage within modern ILW containers. Prior to the commencement of these works the Deflector Plate Assemblies (DPA) and beam structures within the silos are required to be removed to allow access for the waste retrievals equipment through the side of the silo at the point where the DPAs are situated. James Fisher Nuclear (JFN) and Sellafield Ltd have developed a system using suspended water jet cutting equipment combined with custom long reach tooling, installed through penetrations made in the side of the silo, to enable the remote size reduction of the DPA and beams within each compartment. The use of a low pressure suspended water jet cutting system has enabled a single water/abrasive hose to be routed down a tool deployment pole and allowed long hose lengths up to 80 m from pump to nozzle enabling siting of equipment on a congested and difficult access site. The use of lightweight, yet high strength, carbon fibre poles within the tools has enabled the deployment of tool end effectors at distances of up to 6.5 m from the compartment walls using a relatively small 60 mm diameter pole. This maximised articulation of the pole within the 300 mm wall penetrations to reach all areas of the DPA. Capability of the operators to install the tooling and size reduce DPA and beam assemblies have been demonstrated through the training on a Full Scale Test Rig (FSTR). The FSTR was integral to the development of the equipment and techniques and to providing assurance to the client that the project was technically achievable given the tight building restraints and safety issues. (authors)

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DOE OSTI2017

The Micro-Optical Ring Electrode: A Sensor for Multiple Actinide Ions Monitoring - 17586

We present on the development of the Micro-Optical Ring Electrode (MORE) for the analysis of trans-uranium elements in aqueous mixtures. The MORE is a photo-electrochemical device based on a ring microelectrode that uses the insulator interior to the ring as a light guide. This single device exploits the unique photophysical and electrochemical properties of multiple analytes present in mixtures to quantify them. Our study aims to develop a protocol for the analysis of ions of uranium, neptunium and plutonium, which are most relevant to the nuclear industry, especially in the areas of decommissioning and fuel reprocessing, where speedy and safe identification of radioactive contaminants is essential. The technique offers a number of advantages over traditional methods in the area of radiochemical analysis such as real time monitoring in-situ, reduced radiation dose impact to the analyst as a result of greatly reduced sample handling and preparation, improved analysis times allowing quick decision making, at a much reduced cost (e.g. compared to ICP-MS and Scintillation counting). Here we present results of our study on non-radioactive surrogates for the target elements and discuss the advantages the technique can deliver to the nuclear industry. (authors)

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DOE OSTI2017

Drying Simulated Advanced Gas Reactor Fuel: Proving the Concept - 17637

Spent Nuclear Fuel (SNF) reprocessing is due to cease in the UK in 2018 and a new interim storage method may be required if it is found that pond storage is not viable beyond the 25 year period which is currently understood to be safe. Dry storage has been used elsewhere in the world for zircaloy and aluminium clad fuel and has been proposed for use in the UK, however since the UK's fuel is predominantly stainless steel (SS) clad, testing will need to be carried out to ensure that drying methods are suitable. This will require representative samples for macro scale drying tests. To date, little long term stored fuel has been examined. As a result, work has been carried out to try and ascertain whether the corrosion associated with long term storage leads to significant bound water and whether through wall cracks can be formed in intact lengths of SS tube. It was found that unlike an aluminium simulant, corroded SS showed no sign of water loss as a result of oxide dehydration during controlled drying by thermogravimetric analysis, and consequently bound water could be disregarded. It was also found that through wall cracks could be produced in an intact length of SS tube which would be suitable for macro scale drying tests. Some initial macro scale drying tests have been carried out using an alternative test piece with preformed cracks, and three distinct drying regimes to complete dryness have been detected. (authors)

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DOE OSTI2017

Use of advanced optimization algorithms for the design of critical experiments

The need for validating molybdenum and rhodium cross sections in the thermal energy range has arisen for many years for reprocessing activities and burnup credit calculations. An Integral Experimental Request (IER) sheet was submitted by IRSN to the NCSP (Nuclear Criticality Safety Program) in 2015. A preliminary design was therefore proposed for review. In order to define experiments that meet the requirements and that exhibit optimized sensitivities of k{sub eff} to the capture cross sections of {sup 95}Mo and {sup 103}Rh, advanced algorithms of the PROMETHEE workbench, developed at IRSN, were used. The methodology as well as the main results are presented in this paper, which focuses only on the molybdenum experiments design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Regulating Operational Safety at a Future Geological Disposal Facility in the UK - 18038

The United Kingdom (UK) has generated commercial nuclear energy for over 60 years; generating a significant and varied inventory of higher activity radioactive waste (HAW). In accordance with international practice, UK Government policy for long-term management of HAW is disposal in a geological disposal facility (GDF). The nuclear materials comprising HAW include high level waste, intermediate level waste and some low level waste, not suitable for disposal to the Low Level Waste Repository. In addition, other nuclear materials may be reclassified as waste in the future and will require to be managed as HAW for disposal to a GDF; these include spent fuel, uranium and plutonium. The operator of a GDF will be required to obtain a nuclear site licence from the Office for Nuclear Regulation (ONR), as the nuclear safety and security regulator, in order to construct and operate the facility. In the period prior to submission of a licence application, ONR is working to develop its regulatory expectations and provide advice on its regulatory role with regards to a future GDF. ONR has developed regulatory guidance on operational safety, benchmarked against the relevant IAEA Safety Standards and the Western European Nuclear Regulators Association (WENRA) Safety Reference Levels for Disposal. This guidance draws upon ONR's current expectations for nuclear sites; including operational reactors, fuel manufacture and reprocessing facilities, and reactors undergoing decommissioning; and incorporates relevant good practice drawn from international programmes for geological disposal and related industries (eg mining). In addition to the technical assessment guidance, ONR has developed guidance on the licensing process specific to a GDF; providing clarity for example on when a licence will be required. A review of the standard licence conditions applied to all nuclear sites concluded no significant amendment would be required for application to a GDF. Through its voluntary agreement with Radioactive Waste Management Limited (RWM), the developer and future operator of a GDF in the UK, ONR exerts regulatory influence on the design and organisational development of the future licensee. As part of this scrutiny, ONR has assessed RWM's generic Disposal System Safety Case. Subjecting the generic design and safety case to regulatory review provides opportunity for early engagement to ensure the appropriate standards will be achieved. ONR will only grant the necessary regulatory permissions if RWM adequately demonstrates the GDF will be constructed, operated and closed safely. Future challenges faced by ONR with respect to a GDF include building and maintaining regulatory competence and confidence, and engaging effectively with the public as the UK begins its search for a suitable site to provide appropriate assurance that a GDF will be operated safely and securely. (authors)

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DOE OSTI2018

Exploiting Knowledge of Technetium in the Subsurface to Develop Remediation Strategies - 18223

Technetium-99 (Tc) contamination remains a major environmental problem at legacy nuclear reprocessing sites, such as the Hanford Site in south-central Washington State, where the total inventory of released Tc is estimated at 700 curies (Ci). The environmental risk associated with Tc is due to its long half-life (213,000 years) and the high mobility of the oxidized anionic pertechnetate species (Tc(VII)O{sub 4}{sup -}). Thus, there is a critical need to resolve scientific and technical issues underpinning the ability to safely, cost-effectively, and efficiently remediate Tc-contaminated vadose zone sediments and groundwater. Developing enhanced attenuation strategies for released Tc that reduce contaminant migration, while mitigating risk to human health and the environment, requires characterization of any retardation processes and Tc mass flux. The current conceptual model for Tc transport in the vadose zone assumes that Tc is predominantly present in its oxidized form as TcO{sub 4}{sup -}, which is highly soluble and mobile. However, under certain conditions, Tc can become entrained in solid phases. Tc-containing mineral phases from sediment cores extracted from various Hanford waste disposal sites were characterized to determine Tc present in vadose zone sediments as (i) TcO{sub 4} - that remains in water films associated with fine-grained sediments; (ii) Tc co-precipitated with other lanthanides in phosphate mineral phases; and (iii) Tc present as TcO{sub 2}.nH{sub 2}O in localized reducing zones. These results will be used to develop improved long-term Tc remediation strategies that are most likely to succeed during field application, through stimulation of conditions that enhance Tc attenuation. (authors)

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DOE OSTI2018

Electrochemical behavior of Nd and its co-reduction with Bi in molten LiCl-KCl salts - 18360

A molten salt phase is used as the solvent in pyrochemical reprocessing. The molten salt waste generated in pyro-processes contains the residual lanthanides and actinides as various compounds. Various methods have been proposed for recovering and recycling these useful elements. The separation of actinides from lanthanides is performed by electrolysis or reductive extraction. Herein we have studied PyroRedSox-based co-reduction of Nd and Bi via electrolysis in a molten chloride salt. The redox behavior of Nd and the formation potential of intermetallic compounds of Nd and Bi were determined using cyclic voltammetry (CV). Electrorefining was performed under a constant reductive potential at 773 K in LiCl-KCl eutectic mixture; the co-reduction potential was -1.37 V. Lastly, the co-reduction potential of the intermetallic compound was applied to reduce the amount of Nd and Bi. Scanning electron microscopy (SEM) revealed that various kinds of intermetallic compounds were formed in Bi and Nd, and were deposed in the lower part of the molten salt, allowing purification of the contaminated molten salt. (authors)

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DOE OSTI2018

Unique Facilities, Innovative Treatment Processes, and a Talented Workforce Allow Fluor Idaho to Successfully Disposition Transuranic Waste - 18388

Fluor Idaho is utilizing state-of-the-art equipment, treatment technologies, and employee expertise at the Department of Energy's (DOE) Idaho Site to treat contact-handled transuranic (CH-TRU) waste, as well as extremely challenging remote-handled transuranic (RH-TRU) waste forms, to prepare the material for shipment and permanent disposal at the Department's Waste Isolation Pilot Plant (WIPP). Nowhere else in the DOE Complex - perhaps the world - will one find the compilation of tools and an experienced workforce with an unparalleled history of safely retrieving, characterizing, treating, repackaging, certifying, and shipping TRU waste. Since 2003, the Advanced Mixed Waste Treatment Project (AMWTP) has used both conventional and unique retrieval concepts, high-tech characterization equipment - including real-time radiography and radioassay - robotics, and automated treatment processes to retrieve and treat 65,000 cubic meters of above-ground CH-TRU waste. Additionally, engineered controls and specialized personal protective equipment have allowed employees to make thousands of entries into high-radiation areas without a skin or clothing contamination incident - evidence of the employees' commitment to safety. AMWTP has encountered and found solutions for many challenging waste forms, including polychlorinated biphenyl-contaminated wastes, TRU waste containers with up to 1,000 fissile gram equivalent plutonium-239, oversized TRU boxes, and partially absorbed organic sludge waste. By utilizing additional facilities and equipment at the Accelerated Retrieval Project V and VII, Fluor Idaho has developed new processing lines and capabilities to ensure worker safety and milestone completion. These two soft-sided buildings, located at a facility adjacent to AMWTP, were originally constructed to exhume Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) TRU wastes from a section of a 97-acre landfill. After a total of 1,150 cubic meters of targeted waste were removed, personnel, working in collaboration with state regulators, successfully implemented a unique regulatory strategy that allowed these CERCLA facilities to be modified and permitted to treat Resource Conservation and Recovery Act (RCRA) wastes from the AMWTP - a savings of tens of millions of dollars over the cost of constructing new buildings. Adding to the AMWTP's technological uniqueness is its ability to treat wastes from other facilities in the DOE Complex. Two prior decision documents and an agreement with the state of Idaho have allowed AMWTP to treat wastes from 15 other DOE sites. It is the only project in the complex to have done so. At another project located next to AMWTP, targeted waste exhumation within the Subsurface Disposal Area is approximately two years ahead of schedule. Crews have removed more than 8,600 cubic meters of CH-TRU and hazardous wastes from eight exhumation areas that equal a combined footprint of 4.75 acres. Less than one acre remains to be exhumed to satisfy the 2008 Record of Decision waste exhumation end state between the DOE, Environmental Protection Agency, and state of Idaho. Construction of the final building where targeted waste exhumation will continue was completed during the summer of 2017. Exhumation will be completed within the next few years. Following exhumation completion, an evapotranspiration barrier (or cap) will be constructed over the entire landfill. Fluor Idaho continues to make progress on a more challenging project involving an inventory of RH-TRU waste that also contains the reactive metal sodium. Crews are utilizing hot cells, once used for spent nuclear fuel reprocessing and high-level waste treatment at the Idaho Nuclear Technology and Engineering Center, to render this challenging waste form safe for disposal in a national geologic waste repository. The treatment methods being employed on this RH-TRU waste are unique in the DOE Complex. Minute concentrations of water are sprayed onto the sodium-contaminated material in a process called 'spritzing' to chemically change the sodium from a reactive metal to a non-reactive salt. In other cases, sodium-contaminated waste is dipped in water to initiate a reaction in a process called 'immersion.' Lastly, for more challenging waste shapes that could not be successfully treated with the two prior methods, crews lower the sodium-contaminated RH-TRU waste into a one-of-a-kind distillation unit that uses high temperature and vacuum to volatilize the sodium and recapture it in a condenser unit for disposal. The RH-TRU waste treated with this technology is then safe for permanent disposal. Fluor Idaho is proud to operate DOE's CH- and RH-TRU waste treatment facilities, as well as use innovative processes and its talented workforce to complete this important work-scope. Such capabilities allow the contractor to meet the commitments made by the DOE to the state of Idaho. (authors)

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DOE OSTI2018

Delivering Business Benefit to UK Decommissioning through Collaborative Innovation - Approach and Technologies - 18539

The UK's nuclear energy programme, dating from the post-war years, has left a mixed legacy to the country: numerous prototype reactors, fuel-manufacturing plants, research centres, reprocessing plants and 11 power stations. The Sellafield site in west Cumbria houses more than 200 nuclear facilities and 1,000 buildings, making it one of the world's most complex environmental decommissioning challenges. Many 'never-done-before' projects are under way, which require substantial levels of R and D and innovative engineering approaches. One of the Nuclear Decommissioning Authority's (NDA) responsibilities is to ensure that the right level of R and D is carried out to enable progress on delivering the full decommissioning programme. Since 2010, NDA has been collaborating with Innovate UK, the UK government's innovation agency, to engage the supply chain and develop innovative technologies and services to support the NDA's mission and the growing international nuclear decommissioning market. The paper will provide an overview of the approach (e.g. open calls, briefing events, independent assessment, technology showcases) and the associated benefits (e.g. attracting new ideas, end-user engagement, increased supply chain collaboration). The paper also highlights selected technologies developed through the process. (authors)

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DOE OSTI2018

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 18639

Decommissioning of sodium cooled reactors and fast reactor technologies have generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. Battelle Energy Alliance, LLC is evaluating the GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate components from sodium cooled reactors into waste forms that comply with existing disposition pathways. Reactive metal wastes at INL require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics in order to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glassformers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016. The higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results of this two-year demonstration program and future planned activities are presented in the paper. (authors)

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DOE OSTI2019

Unlocking the Vault - Accelerated Delivery Through the Application of Innovation and Collaboration - 19011

The Pile Fuel Cladding Silo (PFCS) at Sellafield was built in 1949/50 to receive and store fuel cladding from reprocessing of Windscale Pile fuel; it has been used to store this material and other waste from the UK nuclear programme ever since. The Silo is unsuitable for indefinite storage; this along with uncertainty in the make-up and condition of the inventory combine to rank the facility alongside the most hazardous in Western Europe. Sellafield Ltd, began working with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS - a joint venture between Bechtel Inc and Cavendish Nuclear Ltd) in 2012, to execute a suite of projects to create an access route to the waste and a capability to retrieve waste from the silo for the first time ever. However, the challenge of providing this capability, while maintaining ongoing safe storage of the waste, and underpinning future performance of both the silo and the waste packages for many hundreds of years, proved to be too much for traditional Sellafield Ltd delivery models. With both cost and schedule forecasts doubling from initial estimates, a new approach was clearly required. The team recognised that success would require all of the experience and expertise of each organisation working together to a common goal. Therefore, significant effort was expended to build a collaborative team environment into which an innovative approach to both the engineering of what was provided and the delivery methods for doing it could be launched. Focussing on collaborative culture in which innovation and fit for purpose thinking were encouraged has completely rebuilt stakeholder confidence. The project is now almost complete; 17 months ahead of schedule and British Pounds 100 m less than the approved cost. (authors)

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DOE OSTI2019

Developing Tomorrows Nuclear Professionals - Apprentice and Graduate Development at PFCS, Sellafield - 19013

The Pile Fuel Cladding Silo (PFCS) at Sellafield, originally built in 1949/50 to receive and store fuel cladding from reprocessing Windscale Pile fuel, has been used to store this material and other waste from the UK nuclear programme ever since. However, the Silo is unsuitable for indefinite storage of the waste; its age, condition and inventory, combined with significant uncertainty in all these factors, rank the facility among the most hazardous in Western Europe. In 2012, Sellafield Ltd began a highly collaborative relationship with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS), to undertake a suite of projects to create a capability for retrieving waste. In doing so, the team has discovered a shared passion for graduate student and apprenticed trades development, which has led to the team goal of becoming the number one destination for training secondments within the associated businesses. This paper describes how the team has leveraged the unique challenge of the project, the extensive experience of the delivery team, and the capabilities of the contributing organisations to create a structure secondment experience aimed at maximizing learning and generating enthusiasm for nuclear clean-up. The scheme was developed and implemented, based on some very simple but consistent feedback from the trainee community on what differentiates a good and bad secondment. The paper includes three case studies from individuals who have participated in the programme and the analysis of overall feedback from all programme trainees. Universally, the scheme has been well received with a significant number of trainees coming back to the project to begin their professional careers. The simple differentiators between a good and bad secondment provide a good basis for those wishing to develop similar schemes. (authors)

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DOE OSTI2019

Revisit of Phase Separation of Borosilicate Glass for Vitrification of High Level Radioactive Waste - 19166

Phase separation of glasses for vitrification of high level radioactive waste has been experimentally studied to know uncertainty of the phase separation conditions; temperature and glass compositions as well as to propose possibility of recycling of heavy metals on a longer time scale. Our current results at 700 and 750 deg. C for ternary borosilicate glass system showed that the phase separation region at 700 deg. C spreads along with a direction of increasing Na{sub 2}O composition when compared with the data in literature. On the other hand, at the glass composition employed in the reprocessing plant in Japan, no segregation was observed. Concerning the application of phase separation phenomena, we had previously reported recovery of almost all elements from the vitrified borosilicate glass containing simulated fission products through phase separation of the borosilicate glass and this suggests that we could propose that after vitrification of high-level radioactive waste heavy metals may be retrieved as energy or materials resources when they are required to utilize on a longer time scale. From this standpoint, phase separation of lead borosilicate glass was also studied to seek economical and versatile recycling. (authors)

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DOE OSTI2019

La Hague STE2 Sludge Retrieval and Conditioning Strategy - 19229

About 9000 m{sup 2} of Long-Lived Intermediate Level Waste LL-ILW STE2 sludges have been generated by the reprocessing operations of UNGG and later UOx fuel in La Hague (UP2-400) and subsequent liquid effluent treatment. They were placed into a set of adjoining tanks, called the STE2 silos. Today, the STE2 Sludges have to be retrieved and conditioned for final disposal in the deep geological repository under development in France (Cigeo). To this end, since the early 2000's, different retrieval and conditioning scenarios have been evaluated. This paper tracks the history of the technical and non-technical key factors having led to the choice of the baseline scenario developed until the end of 2016, describes the technical achievements and makes a state-of the art of the current scenario consisting in retrieving and temporarily storing STE2 sludges and, meanwhile, studying, researching and testing final conditioning processes and technologies in the aim of achieving a technical and economic optimum. (authors)

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DOE OSTI2019

ANSTO Synroc Wasteforms for the Immobilization of a Range of Mo-99 Production Wastes - 19326

Tc-99m, the most commonly used radioisotope in nuclear medicine, is produced from the decay of Mo-99 which is predominantly produced in reactors by irradiation of enriched uranium targets. This results in three waste streams, one of which is a uranium filter cake containing fission products. This filter cake could itself be treated as a waste stream, or processed to retrieve the enriched uranium for further use, thus producing additional waste streams during reprocessing. All wastes are considered intermediate-level wastes and require an appropriate wasteform design for termination. ANSTO is collaborating with NECSA (Nuclear Energy Corporation of South Africa) to develop wasteform and processing solutions for the treatment of intermediate level wastes (ILW) from the production of Mo-99. The work is supported by the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) with technical support from Argonne National Laboratory (ANL). Here we highlight some of the outcomes of this collaboration which has seen the investigation and performance testing of a range of waste treatment technologies. Among the treatment options for the various waste streams, including the enriched uranium filter cake itself, were glass, glass-ceramic, ceramic and geopolymer wasteforms. All wasteforms were characterized (e.g., by X-ray diffraction (XRD) and scanning electron microscopy (SEM)) and their performance characteristics, for example compressive strength and aqueous durability, determined by standard methods and compared. Consideration has been given to the most suitable, practical and cost effective solutions for the immobilization of these Mo-99 wastes. This paper will discuss the down-selection of wasteforms and process technologies for Mo-99 wastes and compare these with regards to wasteform performance and process technology readiness. (authors)

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