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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

What's in WIPP? Packaging TRU Waste to Enhance WIPP's Capacity - 16515

Beginning in 1999, and continuing for 15 years, the US DOE Waste Isolation Pilot Plant (WIPP) repository successfully disposed of defense Transuranic (TRU) waste. Six of the originally planned ten disposal panels were filled before two unrelated accidents temporarily halted disposal operations in February 2014. Based on counting the gross volume of all containers emplaced in WIPP to date, a little more than half of the legislated volume capacity has been subscribed (91,000 cubic meters). This paper describes the many different packaging configurations that have been used to ship TRU waste to WIPP, and makes observations that may enhance the repository capacity when shipments resume. The contents of shipments to WIPP are limited by many requirements of the repository waste acceptance criteria. These criteria derive from three primary sources; 1) the Certificate of Compliance (CoC), issued by the NRC for Type B shipping packaging used by WIPP; 2) The Hazardous Waste Facility Permit, issued by the New Mexico Environment Department; and 3) the WIPP facility's own Documented Safety Analysis (DSA), developed under DOE's regulation 10CFR830. To meet these criteria, various TRU waste forms have been packaged in different ways, with varying packaging efficiency. Of course, there is an overall upper limit on WIPP's volume capacity that was legislated as part of the WIPP Land Withdrawal Act. The volume counted against this limit has been calculated by simply summing the volumes of individual payload containers shipped to and emplaced at WIPP. This method counts significant non-waste volumes inside the containers due to over-packing inner containers and partially filled containers. Without some change to how disposal volumes are calculated, or how waste is packaged, the amount of TRU waste yet to be shipped to WIPP could exceed the legislated capacity. Concepts to enhance the packaging efficiency of future waste shipments to WIPP are presented. Approximate estimates of the waste versus non-waste volumes that have been emplaced to date are discussed. This paper only considers contact handled TRU waste (i.e., contact dose rates < 2 mSv/hr). (authors)

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DOE OSTI2017

Reuse of Ion Exchange Resins for Reduction of Spent Radioactive Resins - 17125

A process is proposed to reuse ion exchange resins. The process consists of ion exchange beds for the ion exchange and regeneration, regeneration solution (regenerant) make-up tanks, an evaporator for concentration of spent regenerant, a dryer for the drying of concentrate, a pulverizer for grinding dried salt and a mixer for polymer solidification of the salt. A spent resins storage tank and a spent regenerant storage tank are also involved in the process. The role of each component in the process for reuse of the resins is explained. In addition, the effect of the reuse of ion exchange resins by regeneration was estimated in view of the reduction of spent resins. As the number of regenerations increases, the volume of spent resins is obviously inversely proportional to the number of regenerations. Although a polymer waste form is generated instead of spent resins when the ion exchange resins are reused, its volume is only about one-fifth of that of spent resins. (authors)

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DOE OSTI2015

Chemical Reactivity and Recommended Remediation Strategy for Los Alamos Remediated Nitrate Salt (RNS) Wastes

Nitrate salt wastes were generated during historical processing operations to recover and purify plutonium at Los Alamos National Laboratory (LANL). These nitrate salts have the potential to behave as oxidizers, and to contain entrained liquids, both of which are prohibited in the Waste Isolation Pilot Plant (WIPP). A process was developed to remediate nitrate salts by mixing salts with an inorganic zeolite pet litter to absorb free liquids, and remove the oxidizing characteristic of the wastes. Process changes introduced instead an organic wheat-­-based pet litter along with organic neutralizers, generating a new waste form referred to as Remediated Nitrate Salt (RNS) waste. The February 2014 radiological release at the WIPP site was traced to a LANL RNS waste drum number 68660. Experimental and modeling studies performed at LANL indicate that mixtures of metal nitrate salts (oxidizer) with Swheat organic kitty litter (fuel) create the potential for exothermic chemical reactions. Even in the presence of triethanol-­-ammonium nitrate (TEAN) created during neutralization, relatively high temperatures, greater than approximately 160 °C (320 °F), are required for runaway exothermic reactions. Evaluation of Drum 68660 characteristics coupled with extensive chemical testing indicate that, in addition to the nitrate salt/Swheat organic kitty litter mixture, an additional trigger mechanism (or mechanisms) is likely required to raise the internal drum temperature high enough to initiate the nitrate salt/Swheat organic kitty litter reaction. A combination of chemical conditions were identified that may lower the temperature for reaction, including initial high acid concentration of free liquids; significant quantities (> 1 gal) of neutralized, absorbed free liquids; the presence of reactive or catalytic metals like magnesium, iron, or lead; the presence of bismuth containing glovebox gloves; and the presence of natural biological activity. Complex surrogate nitrate salt mixtures prepared to simulate wastes, particularly those containing iron and magnesium, can generate NOx gases that partially nitrate the organic Swheat kitty litter and form a more energetic fuel. These complex surrogate salt mixtures display exothermic behavior at temperatures as low as 60 °C (140 °F) which is still well above the ambient temperature conditions experience by a drum. Neutralization of free liquids, and sorption onto Swheat establishes conditions (moisture) with near-neutral pH) that will support natural biological activity. Spontaneous self-heating generated by low-level chemical reactions and/or the respiration of bacteria, molds, and microorganisms is potentially important in the early stages and may be sufficient to raise the temperature as high as 60 °C (140 °F), where the other exothermic chemical reactions can take place. Additional studies are being conducted to evaluate the role biological activity may have played in initiating the event. The use of Swheat absorbent in the processing of nitrate salt wastes can be pinpointed as the critical processing decision that led to the failure of drum 68660, regardless of the ii details of the thermal processes that enabled the drum to achieve temperatures sufficient to initiate the chemical reactions.

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DOE OSTI2013

The benefits of a fast reactor closed fuel cycle in the UK

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 19092

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. (authors)

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DOE OSTI2013

Building Confidence in LLW Performance Assessments - 13386

The performance assessment process and incorporated input assumptions for four active and one planned DOE disposal sites were analyzed using a systems approach. The sites selected were the Savannah River E-Area Slit and Engineered Trenches, Hanford Integrated Disposal Facility, Idaho Radioactive Waste Management Complex, Oak Ridge Environmental Management Waste Management Facility, and Nevada National Security Site Area 5. Each disposal facility evaluation incorporated three overall system components (1) site characteristics (climate, geology, geochemistry, etc.), (2) waste properties (waste form and package), and (3) engineered barrier designs (cover system, liner system). Site conceptual models were also analyzed to identity the main risk drivers and risk insights controlling performance for each disposal facility. (authors)

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DOE OSTI2018

Decommissioning Concept of Nuclear Fuel Cycle Facility - 18255

As the Korean Nuclear Safety Act was revised, it is urgent to prepare a decommissioning plan for the nuclear fuel cycle facilities operated by the Korea Atomic Energy Research Institute, KAERI. First, the basic concepts necessary for dismantling each facility and the order of dismantling for each dismantling work were determined. In this paper, we consider the disassembly order of the nuclear fuel cycle facilities such as the post irradiation fuel examination facility, the research reactor fuel production facility, the radioactive waste treatment facility, the radioactive waste form examination facility. The decommissioning plan for the research nuclear fuel cycle facilities will be based on the abovementioned dismantling basic concept and disassembly sequences. (authors)

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DOE OSTI2018

Simulating the Effect of Fracture Connectivity on Repository Performance with GDSA Framework - 18589

This work examines the possible migration of radionuclides from a deep geologic nuclear waste repository sited in fractured crystalline host rock. The key safety concern is the potential effect on waste isolation of the inter-connectivity of the fracture network, which is primarily established by the temporal evolution of the temperature and stress fields at the time of original rock intrusion and cooling. Two end members are considered in these simulations, one with a high degree of connectivity to the biosphere, such that advective transport through the fracture and fault network controls radionuclide migration, and the other with a low degree of connectivity, such that slow diffusive transport through the crystalline rock matrix is the controlling process for radionuclide migration to the biosphere. Both end members have several areally extensive, high-transmissivity deformation zones, whose distance to the repository is an important factor for waste isolation capability. Uncertainties in fracture properties (transmissivity, orientation, radius) and fracture distribution also give rise to uncertainty in waste isolation capability (i.e., the overall connectivity of the repository with the surface biosphere). These types of natural system uncertainties are generally present in any geologic site-characterization program and thus represent an important factor in assessing repository performance in hard rock environments. The waste considered in these simulations are spent fuel rod assemblies from the U.S. commercial nuclear reactor fleet. Thermal output of the waste must be considered in the simulations and can influence the rate and timing of waste package failure, waste form degradation, and fluid flux due to thermal expansion around the repository horizon. To achieve a high-fidelity representation of radionuclide transport in fractures and rock matrix, combined with thermal energy transport and fluid flow in fractures and matrix, the mathematical model is solved numerically in a parallel high-performance computing (HPC) environment on a finite volume unstructured grid consisting of approximately 5 million cells, using Geologic Disposal Safety Assessment (GDSA) Framework (https://pa.sandia.gov), an open-source performance assessment tool for deep underground disposal of nuclear waste. GDSA Framework uses PFLOTRAN to solve the balance equations on a three-dimensional grid with heterogeneous properties, using multiple processors in a parallel configuration based on domain decomposition. The fracture networks in these simulations are originally generated as discrete fracture networks (DFNs), which are sets of two-dimensional planes distributed in a three-dimensional domain. The method used in GDSA Framework maps the stochastically generated DFN to an equivalent continuous porous medium (ECPM) domain that allows for the simulation of coupled heat flow, fluid flow, and radionuclide transport, including heat conduction through the matrix of the fractured rock, which is a process not easily modeled in a DFN representation. Computational efficiency is also greatly enhanced using the ECPM method, allowing for a realistic representation and analysis of uncertainties in a multi-realization performance assessment of a deep geologic repository. The effect of fracture connectivity on the waste isolation safety function, as brought to light by these GDSA Framework simulations, points to the importance of including a realistic representation of uncertainties in fracture properties and distribution (effectively, uncertainty in spatial heterogeneity) in repository safety assessment simulations. (authors)

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DOE OSTI2012

Carbonation Behavior of Pure Cement Hydrates under Supercritical Carbon Dioxide Conditions - 12199

Carbonation of cement-based waste forms using a supercritical carbon dioxide (SCCO{sub 2}) is a developing technology for the waste immobilization of radioactive and non-radioactive wastes. However, the detail carbonation behaviors of cement matrices under the SCCO{sub 2} condition are unknown, since cement matrices forms very complex phases. In this study, in order to clarify the crystal phases, we synthesized pure cement hydrate phases as each single phases; portlandite (Ca(OH){sub 2}), ettringite (Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26H{sub 2}O), and calcium silicate hydrate (n CaO---m SiO{sub 2} ---x H{sub 2}O), using suspensions containing a stoichiometric mixture of chemical regents, and performed carbonation experiments using an autoclave under supercritical condition for carbon dioxide. The XRD results revealed both the carbonate phases and co-product phases depending on the initial hydrate phases; gypsum for Ettringite, amorphous or crystalline silica for calcium silicate hydroxide. Thermogravimetric analysis was also performed to understand carbonation behaviors quantitatively. According to the experimental results, it was found that the major reaction was formation of calcium carbonate (CaCO{sub 3}) in all cases. However, the behaviors of H{sub 2}O and CO{sub 2} content were quietly different: Portlandite was most reactive for carbonation under SCCO{sub 2} conditions, and the CO{sub 2} content per one molar CaO was ranged from 0.96 ∼ 0.98. In the case of Ettringite, the experiment indicates partial decomposition of ettringite phase during carbonation. Ettringite was comparatively stable even under the SCCO{sub 2} conditions. Therefore, a part of ettringite remained and formed similar phases after the ettringite carbonation. The CO{sub 2} content for ettringite showed almost constant values around 0.86 ∼ 0.87. In the case of calcium silicate hydrate, the carbonation behavior was significantly influenced by the condition of SCCO{sub 2}. The CO{sub 2} content for the calcium silicate hydrate had values that ranged from 0.51 ∼ 1.01. The co-products of the carbonation were gypsum (CaSO{sub 4}) for ettringite, silica gel (SiO{sub x}) and silica (SiO{sub 2}) for calcium silicate hydrate, which also contributed to the densification of the particles. The production of co-products enhanced the change to their morphology after the carbonation. (authors)

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DOE OSTI2019

The Next Generation System Analysis Model: Capabilities for Simulating a Waste Management System - 19131

The U.S. Department of Energy (DOE) Office of Nuclear Energy Office of Spent Fuel and Waste Disposition's Integrated Waste Management System (IWMS) program is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWMS are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The system architecture analyses consider alternative disposition pathways for commercial spent nuclear fuel (SNF) being stored dry in at-reactor Independent Spent Fuel Storage Installations (ISFSIs) or wet in spent-fuel pools. One method for evaluating these alternatives is simulating the movement of SNF throughout the system components. Successfully simulating a system on the scale of a national waste management system requires - accurate baseline information (e.g., historic fuel discharges from reactors and dry storage loadings at reactor sites), - modeling all logistic interactions between components of the proposed system, and - tracking and reporting all relevant details regarding the simulation's execution. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWMS. It is based on the Performance Analysis Tool, a simulation platform that has been used to model logistics for the Department of Defense and the Federal Emergency Management Agency. Using existing, government-owned software allows NGSAM developers to modify the software at the level of source code and to develop flexible usability and modeling solutions for developing and running simulations of the IWMS. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. NGSAM is designed with the flexibility to model and track the behavior of virtually any IWMS architecture assumptions or design considerations. For example, NGSAM can model multiple interim storage facilities or repositories or none at all. Similarly, all assumptions regarding shipment schedules, operating schedules, and reactor operation assumptions can be modeled and adjusted in NGSAM with relative ease. The development of NGSAM was targeted to waste management system analysts as the primary users. With that in mind, analysts from ORNL, Sandia National Laboratories (SNL), and Argonne National Laboratory (ANL) were intimately involved in the development process. That involvement extended beyond merely defining requirements. System analysts have been involved with design decisions and software benchmarking and tests, and they have provided subject-matter expertise where necessary. The result is a high degree of interaction between professional software developers and system analysts, with a multitude of benefits to both groups. This paper explores some of the underlying methodology of the agile software development process used in the creation of NGSAM. It illustrates the capabilities of NGSAM. Some example use cases are discussed as well, with relevant results providing context for how NGSAM is used by system analysts to gain a greater understanding of alternative waste management system architectures. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed-to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Enhanced Performance Assessment Models for Generic Deep Geologic Repositories for High-Level Waste and Spent Nuclear Fuel - 16223

Development of an enhanced performance assessment (PA) capability for geologic disposal of spent nuclear fuel and high-level waste has been ongoing for several years in the U.S. repository program. The new Generic Disposal System Analysis (GDSA) modeling and software framework is intended to be flexible enough to evolve through the various phases of repository activities, beginning with generic PA activities in the current Concept Evaluation phase to site-specific PA modeling in the Repository Development phase. The GDSA Framework utilizes modern software and hardware capabilities by being based on open-source software architecture and being configured to run in a massively parallel, high-performance computing (HPC) environment. It consists of two main components, the open-source Dakota uncertainty sampling and analysis software and the PFLOTRAN reactive multi-phase flow and transport simulator. Reference cases or 'generic repositories' have been, and are being developed, based on typical properties for potential salt, clay, and granite host-rock formations and corresponding engineered design concepts for each medium. Past simulations have focused on a generic repository in bedded-salt host rock, while the most recent research has focused on a reference case for a typical clay/shale host rock. A variety of single-realization (i.e., deterministic) and multi-realization (probabilistic) results for the new clay reference case are presented, including an analysis of the effects of heat generation on repository performance, assuming a 100-year out-of-reactor commercial SNF waste form. Order-of-magnitude differences between predicted radionuclide concentrations in thermal versus isothermal simulations imply that mechanistic, coupled-process modeling in three-dimensional (3-D) domains can be important for building confidence in post-closure performance assessments. (authors)

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DOE OSTI2020

Criticality Consideration for Geologic Disposal of DOE-Owned Spent Nuclear Fuels - 20255

Decisions must be made relative to packaging spent nuclear fuel (SNF) for storage, transportation, and eventual disposal. This paper evaluates potential criticality-related loading and packaging considerations for Department of Energy (DOE)-owned SNF that could be impacted favorably or unfavorably by disposal uncertainties and proposes approaches to minimize such impacts. Using Al-Based SNF as an example and drawing on analyses performed in support of the Yucca Mountain Repository License Application (LA), this paper explores the role of natural barriers, engineered barriers, and waste form characteristics relative to reducing the potential for postclosure criticality. Several parameters important to postclosure criticality are considered in order to determine which are useful discriminators to support packaging decisions. Using these parameters, a qualitative assessment is performed relative to three options for managing DOE-owned SNFs. (authors)

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DOE OSTI2016

Deep Borehole Disposal Safety Analysis.

This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

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DOE OSTI2016

The Characterisation and Removal of Water Droplets in High Pressure Water Jetting Nuclear Decontamination - 16036

The use of high pressure water jetting as a decontamination method has seen limited deployment within decommissioning operations in the UK. The full extent of secondary waste forms produced as a result of such operation has not been previously investigated. The aim of this investigation was to determine the process of water droplet formation within the decontamination area and subsequent air treatment pathway. Evaluating the potential security hazards associated with the most widely implemented air filters used as a radiological protection barrier. (authors)

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DOE OSTI2018

Breaking Through Historical Biases in Performance Assessments - 18056

Performance assessments (referred to as safety assessments internationally), which include mathematical modeling of natural and engineered systems for many hundreds and thousands of years, play an important role in decision-making for disposal of radioactive waste and cleanup of sites and facilities contaminated with radioactive material. These decisions can involve options costing hundreds of millions or even billions of dollars. Thus, it is critical to understand the underlying assumptions behind regulatory and technical aspects of these assessments to better understand and communicate the basis for decisions. Policy and technical biases deliberately built into early assessment approaches allowed safety margin to account for uncertainties known to be present, but not quantitatively assessed. Many of these biases (e.g., regulatory standards, human habits, effectiveness of waste forms, inadvertent intrusion assumptions, etc.) continue to exist in spite of advances in technical approaches and knowledge. These built-in biases, often not effectively communicated, can lead to misleading interpretation of results and potentially influence public confidence and costly regulatory decisions regarding the safety of a disposal facility or a closure action. (authors)

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DOE OSTI2020

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 20466

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by the Department of Energy (DOE or Department). No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and a lack of Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository. (authors)

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DOE OSTI2012

Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass-ceramic waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1; • Feeding, melting, and pouring: TRL-1; • Glass ceramic formulation: TRL-1; • Canister cooling and crystallization: TRL-1; and • Canister decontamination: TRL-4. Although the TRL’s are low for most of these CTE’s (TRL-1) primarily because the specific waste stream is not known or tested, the effort required to advance them to higher values is relatively low. A TRL of 2 would be obtained by completing an initial waste composition/property estimate, a preliminary engineering study, some additional laboratory scale tests of the glass-ceramic, and a mixing and sampling test. Relatively little additional effort is required to advance the technology to TRL-3.

Glass ceramics↗
DOE OSTI2020

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗