THE UNITED STATES PERSPECTIVE ON POST-CLOSURE CRITICALITY ASSESSMENTS IN THE FINAL DISPOSAL OF HIGH-LEVEL WASTE
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Price, Laura L.↗
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Product Consistency Test Results for the HLW HAlG Glasses
This report summarizes the chemical analysis of Product Consistency Test leachates received from Pacific Northwest National Laboratory (PNNL). The leachates are from a series of quenched simulated nuclear waste glasses designated High-Level Waste High Aluminum Glass that were designed and fabricated at PNNL. The reported data is provided to be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The elemental release for the study glasses is reported as normalized concentration NCi. NCi of several elements was computed for both the target and measured glass compositions. Several of the glasses exhibited NC B , NC Li , and/or NC Na values that were greater than the Environmental Assessment (EA) benchmark values. Several of the leachates from approved reference material (ARM) glasses included with the study glasses had elemental concentrations greater than the expected ranges. Two water blanks contained measurable amounts of Si; measurements confirmed by rerun samples. One water blank also contained a measurable amount of Na.
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Characterization of the Sulfur-Saturated Melt Versions of the HLW HAlG Glasses
This report provides the results from the chemical analyses of a series of sulfur-saturated melt versions of the High-Level Waste High-Aluminum Glass study glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the sulfur-saturated melt versions of the glasses to allow for comparisons with targeted compositions as well as the measured compositions of the quenched glasses. The relative differences between the targeted and measured concentrations of B 2 O 3 , Cr 2 O 3 , K 2 O, Li 2 O, Na 2 O, and P 2 O 5 for several of the glasses were greater than ±10%. The relative differences between the targeted and measured concentrations of Al 2 O 3 and ZrO 2 for one of the glasses were greater than ±10%. As expected, the measured concentrations of SO 3 in most of the glasses were higher than targeted due to the use of the sulfur saturation method in fabricating these glasses. The wash solutions contained mainly sodium, sulfur, and sulfate ions.
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Dissolution of Fluoride Salts in Hanford Tank Waste
The Direct Feed High-Level Waste (DFHLW) strategy seeks to bypass the Hanford Waste Treatment and Immobilization Plant Pretreatment Facility while retaining some processing functions to maximize waste feed loading and minimize high-level waste (HLW) waste volume. The DFHLW flowsheet needs leaching, washing, and solids concentration operations either in new or existing tanks. The effectiveness and efficiency of sludge washing has a substantial impact on DST space, mission duration, and the evaporation and low-activity waste (LAW) treatment operations required by these large wash-water additions. One target species requiring washing is fluoride. The HLW glass composition limits for fluorine drive operations to dissolve fluoride-bearing salts into the LAW fraction and thereby maximize waste loading in HLW glass. The fluoride in many high-level wastes at Hanford is predominantly in the form of fluoride-salt precipitates: villiaumite (NaF), kogarkoite (Na 3 FSO 4 ), and natrophosphate (Na 7 F(PO 4 ) 2 ·19H 2 O). Fluoride produces melter off-gas that creates corrosion risk in the off-gas system piping, while the sulfate and phosphate in the fluoride double salts kogarkoite and natrophosphate can be detrimental to glass waste loading. Fluoride salts are sparingly soluble, with solubilities ranging from approximately 40 to 130 kg per kL of pure water, and the dissolution kinetics of the three fluoride salts are not well known. Unexpected delays in a tank dissolution process could be encountered as a result of the lack of information about dissolution rates. In addition, if the double salts show transient non-stoichiometric dissolution of fluoride versus phosphate or sulfate, unexpectedly high concentrations of one of these other constituents could be produced. Washington River Protection Solutions authorized Pacific Northwest National Laboratory to collect the available data for fluoride salt dissolution rate, provide a scoping estimate of dissolution time if possible, and identify gaps in the understanding and predictive capability for estimating dissolution time. Open literature and Hanford reports were reviewed to document, understand, and (where possible) evaluate limitations on fluoride salt equilibria and dissolution kinetics, including both mass transport and surface reaction rate. Scoping estimates of dissolution time were made for mass-transfer-controlled dissolution of spherical particles of fluoride salts suspended in liquid. This is not the only potential governing mechanism; dissolution could be substantially slower if the surface reaction rate (the rate of release of ions from the surface) is the controlling mechanism. When the minimum amount of water for complete dissolution is used and the slip velocity between the liquid and suspended particles is less than or equal to the terminal settling velocity, the estimated mass-transfer rates allow 0.1 mm particles of fluoride salt to dissolve in minutes at 25 °C in water containing no other dissolved salts. Much larger solids, such as the 6-mm chunks that have been seen in heels, could take a few hours to more than a week to dissolve. The actual dissolution times will depend strongly on the actual slip velocity, the extent of particle suspension, constraint by surface reaction rates, the ratio of solvent to solid, and the presence of common ions that shift the solubility equilibria to restrict dissolution of fluoride salts.
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Phosphate and Fluoride Processing Options for Hanford Sludge
The Direct Feed High-Level Waste (DFHLW) strategy represents an alternative flowsheet to bypass the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Facility. Key processing functions previously captured during baseline pretreatment operations include leaching and washing prior to solids concentration and should be considered in the new DFHLW flowsheet to maximize waste feed loading, minimize high-level waste (HLW) volume, and mitigate corrosion challenges associated with vitrification of high phosphate and fluoride concentrations. The effectiveness and efficiency of sludge washing can have a substantial impact on DST space and mission duration. Two target species that benefit significantly from washing are phosphate and fluoride. Commonly present in many high-level wastes at Hanford, phosphate and fluoride are found predominately in the form of salt precipitates and can have adverse and detrimental effects on glass waste loading. Additionally, fluoride produces melter off-gas that creates corrosion risks in the off-gas system piping. Determining the solubilities of fluoride and phosphate salts, as well as the dissolution kinetics under potential HLW processing conditions will help to mitigate future operational risks and mature flowsheet technical bases.
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Panel Session 110B: US DOE High-Level Radioactive Waste (HLW) Interpretation
This panel focused on the US DOE High-Level radioactive Waste (HLW) definition interpretation. Discussion integrated topics such as the issuance of the HLW interpretation Federal Registration Notice, the status of the draft Environmental Assessment that analyzes the treatment and commercial disposal of up to 10,000-gallons of SRS Defense Waste Processing Facility recycle wastewater in South Carolina, and the path forward. Panelists with presentations: Status on the Environmental Assessment for the Commercial Disposal of Defense Waste Processing Facility Recycle Wastewater from SRS (Theresa Kliczewski); Enhancing Stakeholder Engagement on the HLW interpretation (Kara Colton, Rick McLeod); Technical Aspects of Potential Disposal of SRS DWPF Recycle Wastewater (Kent Rosenberger)
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Ruthenium solubility and its impact on the crystallization behavior and electrical conductivity of MoO 3 -containing borosilicate-based model high-level nuclear waste glasses
The present study focuses on investigating the solubility of RuO 2 in a borosilicate-based model high-level waste glass and understanding its impact on the crystallization behavior and electrical conductivity of the resulting vitrified waste forms. The solubility limit of RuO 2 in the investigated glass composition has been determined to be 460 ppm by weight. The higher concentration of RuO 2 results in sub-micron sized Ru-rich inclusions in the glassy matrix, which eventually agglomerate to form needle-like and polyhedral RuO 2 crystals. It is observed that RuO 2 selectively promotes the crystallization of the rare-earth apatite phase over the powellite phase. The as-synthesized RuO 2 -containing glasses exhibit semiconducting behavior with a similar level of electrical conductivity below the glass transition. Here, the percolation of non-uniformly distributed RuO 2 inclusions may result in a formation of short-range low-resistive conduction pathways in the host glass matrix leading to an apparent metallic-like behavior of selected thin samples with the highest ruthenium content.
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Surface interfacial analysis of simulant high level nuclear waste glass dissolved in synthetic cement solutions
Abstract The corrosion mechanisms and kinetics of a Mg-rich alkali aluminoborosilicate glass simulating UK high-level waste (CaZn28) were investigated upon dissolution in synthetic cement solutions. Dissolution varied as a function the different pH and alkali/alkaline earth content of each cement solution. High resolution microscopy and spectroscopy techniques ascertained the nature of the interface between the glass and the cement solutions. TEM-EDS revealed alkali- and alkaline earth-rich silica gels, into which K, Ca and Mg were incorporated. TEM-SAED, combined with synchrotron micro-focus XRD, identified the ubiquitous precipitation of the Mg-aluminate layered double hydroxide phase, meixnerite (Mg 6 Al 2 (OH) 18 ·4H 2 O), in addition to goethite (FeOOH) and crystalline silica. The C-S-H phase, tobermorite (Ca 5 Si 6 O 16 (OH) 2 ·4H 2 O), was identified in the most Ca-rich solution only. These data give insight to the role of alkali/alkaline earth-rich solutions in the dissolution or radioactive waste glasses, of importance to the final disposition in a geological disposal facility.
Corkhill, Claire L. (ORCID:0000000274883219)↗
Overview of Salt Repository Research and Development for Spent Nuclear Fuel and High-level Nuclear Waste in the United States - 20307
This paper summarizes the current US Department of Energy Office of Nuclear Energy's (DOE-NE) work towards developing and executing a research and development program that addresses both scientific and technical issues related to long-term disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a hypothetical bedded salt based geological repository. A primary goal of the program is to create a generic Geologic Disposal Safety Assessment (GDSA) Framework that can be used to help guide decisions on siting a possible future bedded salt repository. The generic GDSA work includes analysis of the impacts of heat generation caused by decay of short-lived radionuclides. We report progress in four primary areas. First, we discuss the development and recent modifications of a research and development road-map. Second, we briefly describe an experimental approach to better understand thermal processes in salt. Third, we highlight collaborations with the international research community that leverage salt-based repository science around the world. Finally, we discuss how our findings are being used to aid in the development of a generic safety assessment for a bedded salt repository containing SNF and HLW. (authors)
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Results of Re-evaluation of FEPs Related to Implementing the ABD Glass Program
The Savannah River Site plans to reprocess defense spent nuclear fuel currently stored in their L-Basin via the Accelerated Basin Deinventory (ABD) Program. The previous plan for the L-Basin spent nuclear fuel was to dispose of it directly in the federal repository without reprocessing. Implementing the ABD Program will result in final disposal of approximately 900 fewer canisters of defense spent nuclear fuel and the production of approximately 521 more canisters of vitrified high-level waste glass with some specific differences from the planned high-level waste glass. Because the 235U in the L-Basin spent nuclear fuel is not intended to be recovered, the fissile mass loading of the vitrified high-level glass waste form to be produced must be increased above the current value of 897 g/m 3 to a maximum of 2,500 g/m 3 . Therefore, implementing the ABD Program would produce a variant of high-level waste glass—the ABD glass—that needs to be evaluated for future repository licensing, which includes both preclosure safety and postclosure performance. This report describes the approach to and summarizes the results of an evaluation of the potential effects of implementing the ABD Program at the Savannah River Site on the technical basis for future repository licensing for a generic repository that is similar to Yucca Mountain and for one that is fully generic. This evaluation includes the effects on preclosure safety analyses and postclosure performance assessment for both repository settings. The license application for the proposed Yucca Mountain repository (DOE 2008), which is serving as a framework for this evaluation, concluded that the proposed Yucca Mountain repository would meet all applicable regulatory requirements. The evaluation documented in this report found that implementing the ABD Program is not expected to change that conclusion for a generic repository similar to Yucca Mountain or for a generic repository with respect to the preclosure safety analyses. With respect to the postclosure performance of a generic repository, no concerns were identified.
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Update on the Status of Deep Borehole Disposal of High-Level Radioactive Waste in Germany - 20285
The phase-out of nuclear energy in Germany will take place in 2022. A site for final disposal of high level radioactive waste (HLRW) has not yet been found, but a site selection process was restarted by Act on the Search for and Selection of a Site for a Disposal Facility for High-Level Radioactive Waste (Site Selection Act - StandAG 2017). This act was based on a recommendation by a commission which also advised to follow up the development of deep borehole disposal (DBD) as an alternative option for final disposal of HLRW. This paper summarizes briefly the status of DBD in Germany and if this option should be pursued in Germany. Although there are some merits of DBD, it can only be a real option if research and development is supported. The technical equipment for boreholes of the required size will only be developed if there is funding and a feasibility test. Furthermore, any concept of DBD and technology must be detailed further, and some requirements of the act must be reconsidered. The paper concludes that despite the possible merits of DBD, the political and financial support for R and D will only be provided for the time being if DBD is pushed by interested parties. Alternatively, if the site selection procedure is not progressing well, then the alternative option of DBD may find greater interest in the future. However, its availability will be severely limited due to the missing development of detailed concepts and its demonstration of feasibility. (authors)
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System Configuration Evaluation for Process Settling of Hanford Waste Solid Particles
Direct Feed High-Level Waste (DFHLW) is a potential flowsheet operations approach to initiating high-level waste (HLW) vitrification prior to completion of the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Facility. A settle/decant process has been proposed to concentrate solids prior to delivery to the WTP HLW Facility during DFHLW operations, wherein the solids in a settled layer would be remixed with the supernatant liquid remaining after decanting operations to provide the feed at required solids concentrations. Settling would be used in lieu of purpose-built filtration or other solids separation equipment. Pacific Northwest National Laboratory (PNNL) is providing baseline technical support to the Washington River Protection Solutions (WRPS) Flowsheet Integration group. To support planning for DFHLW, WRPS previously requested that PNNL evaluate the current data set available to predict the time needed for HLW solids to settle and the solids concentration and strength of that settled layer, to identify gaps in the understanding and predictive capability of HLW solids waste settling times, and to provide scoping estimates of the potential settling times. Eight technical gaps were identified for predicting settling times and characteristics of the formed sediment layers. In addition to the data gaps, an overarching observation was made that there is significant variation in behavior of settling rate and settled layer data. The settling time required to concentrate solids via a settle/decant process was determined from the limited data to have a difference of potentially more than a factor of 5,000 in the estimated settling times, varying from 0.2 to 1,060 days for example depending on process vessel depth and final sediment solids concentration. In contrast, successful processes of liquid-forward output streams resulting from in-tank settling and decanting forward liquid have been reported for operations conducted at the Hanford Site. The purpose of this current report is to further support DFHLW planning by evaluating double-shell tank (DST) and alternate vessel equipment and operational configurations to enable optimization of the settle/decant process to concentrate solids. Hanford waste processing behavior specific to liquid feed availability following a slurry transfer in a DST is summarized, including process stream characteristics and process equipment configurations. The performance of DST process equipment configurations is evaluated for possible improvements using computational fluid dynamics (CFD) and simple analytical models. Potential new vessel design(s) specific to enabling effective settle/decant processes, and cursory summary of other separate and inline solids separations processes, are also provided. The CFD results indicated that improvement in outflow solids concentration was promoted by a reduction in the slurry flow rate, angling the distributor nozzles downward, and lifting the transfer pump. The solid-liquid analysis evaluating particle trajectory confirmed that the potential for particle ingestion (in the transfer pump) was decreased with increased radial separation between the inlet and outlet (transfer pump inlet), decreased inlet flow, and decreased liquid density and viscosity for a neutrally buoyant inlet flow. An assessment was also made of the potential for inflow configuration changes to result in the discrete mounding or piling of solids within the tank. Based on the characterization of the settled waste to date, HLW sediments will be unlikely to sustain a substantial angle of repose to facilitate significant variations in the elevation of the settled solids.
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Savannah River Site High Level Radioactive Waste Tank Sample Thermolytic Hydrogen Generation - 20346
To support the revision of flammability calculations and controls for the Savannah River Site (SRS) High-Level Radioactive Waste (HLW) tanks, Savannah River National Laboratory (SRNL) conducted laboratory measurements with the goal of quantifying hydrogen produced in HLW by non-radiolytic chemical reactions (i.e., thermolysis). Testing was performed with HLW tank supernate samples that represented a cross section of types of waste at SRS, including tanks that store typical evaporator concentrate, dissolved salt-cake, dilute recycle stream from the Defense Waste Processing Facility (DWPF), evaporator concentrate of the recycle stream from DWPF, fresh waste from the canyon separations facility, and waste that has been through salt processing cesium removal. Non-radioactive simulant testing included the classes of organic compounds historically introduced into the HLW tanks that were recently shown to be the most active toward the thermolytic generation of hydrogen. The reaction rate equations developed for thermolytic hydrogen generation from each class of organic compound showed direct proportionality to organic compound concentration and hydroxide concentration. Simulant testing identified that highly concentrated waste with high hydroxide concentration had the highest rates of thermolytic hydrogen generation. Using the radioactive tank sample thermolysis measurements at high temperatures and the mechanistic salt dependence of the simulant models, a global model was developed for tank waste thermolytic hydrogen generation as a function of total organic carbon content of SRS HLW. Through the primary functionality of the global model (organic carbon and free hydroxide concentration), the relative reactivity of the organic carbon in the waste was quantified. (authors)
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Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241
The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)
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Review of corrosion interactions between different materials relevant to disposal of high-level nuclear waste
Abstract This review covers the corrosion interactions between different materials that are relevant to the disposal of high-level nuclear waste, in particular the waste forms and containers. The materials of interest are borosilicate glass, crystalline ceramics, metal alloys, and any corrosion products that might form. The available data show that these interactions depend on the structure, chemistry, thermodynamic history, and proximity of the materials in contact, as well as the environmental attributes, such as temperature, solution chemistry, and radiation. Several key mechanisms that govern these interactions are highlighted. Scientific gaps and open questions are summarized and discussed.
Materials Science↗
An analysis of the technical status of high level radioactive waste and spent fuel management systems
The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.
English, T.↗
Hydro-mechanical behavior of heated bentonite buffer for geologic disposal of high-level radioactive waste: A bench-scale X-ray computed tomography investigation
The behavior of heated bentonite buffer is critical for the security and long-term performance of a geological repository for high-level radioactive waste (HLW). While laboratory column experiments have been conducted to investigate compacted bentonite and coupled THMC (thermal-hydro-mechanical and chemical) processes for a moderate temperature range of up to 100 °C, data for a higher temperature range are limited. Understanding bentonite behavior and coupled THMC processes under higher temperatures (e.g., up to 200 °C) could allow for a more economic repository design and would expand the data and knowledge base for more reliable modeling. In this study, a bench-scale experiment was conducted in a compacted bentonite column experiencing both heating up to 200 °C in the center and hydration from a sand-clay boundary surrounding the column. During the experiment run for 1.5 years, frequent X-ray computed tomography (CT) scanning of bentonite provided insights into the spatiotemporal evolution of (1) hydration/dehydration, (2) clay swelling/shrinkage, (3) displacement, and (4) mineral precipitation. After the experiment, a comprehensive post-dismantling characterization of bentonite samples was conducted. Results showed that the bentonite hydration was axi-symmetrical despite the initial heterogeneity due to packing, confirming the ability of bentonite to seal fast flow/transport paths. Further, compared to a non-heated control experiment, the heated column showed greater CT density variations along the radial distance, indicating that homogenization of bentonite might be more difficult if a temperature gradient is maintained in the repository. Precipitation of an anhydrite layer occurred in the inner hot zone, pointing to potential concerns about salt precipitation causing canister corrosion. Ultimately, the experiments provided a high-resolution window into the strongly dynamic and coupled behavior of bentonite exposed to heating, hydration and swelling, which will be valuable for improving modeling of coupled processes, especially for the early state of a HLW repository.
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Composition Measurements of the HLW HAlG Glasses
This report provides the results from the chemical analyses of the glass compositions of the High-Level Waste High-Aluminum Glass study glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with targeted compositions. The relative differences between the targeted and measured concentrations of Cr 2 O3, P 2 O 5 , and ZrO 2 for several of the glasses were greater than 10%. The relative difference between the targeted and measured concentrations of CaO was greater than 10% for one glass. The relative difference between the targeted and measured concentrations of Li 2 O was greater than 10% for one glass. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.
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