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DOE OSTI2019

Optimization of Backfill Material Transport for the Future Belgian Geological Disposal Facility - 19071

In Belgium, long-lived, low- and intermediate level waste (B waste) will be conditioned in concrete monoliths B and high-level waste (C waste) in so-called Supercontainers (SC). It is planned to dispose of the waste packages in a Geological Disposal Facility (GDF). The B and C waste shall be disposed of each in one emplacement field at a depth between 200 m and 600 m below the surface. Two access galleries run through each field and connect the disposal galleries with a central shaft and support zone. After the emplacement of a specified number of waste packages, a stepwise backfilling of the voids is planned in order to stabilize the galleries, to limit the amount of voids, and to ensure a favourable chemical environment. Taking into account the objectives of backfilling and the specific material properties of the waste packages two mortars were developed to backfill the disposal galleries. In particular, for reasons of safety at work, the mortars should be hydraulically pumped through pipelines. Accordingly, in any case, it is necessary to operate storage facilities of the raw materials, a mix plant and a pump. The raw materials are foreseen to be supplied by trucks from external producers. Due to this fact, the better accessibility of the plants and with the aim of minimizing noise and dust emissions and the space for technical systems in the GDF, it is advisable to install the technical equipment above ground outside the GDF control area. However, the feasibility of this basic concept depends on the processing time of the backfill materials and the capacity of the transport system. Thus, although short transport times can theoretically be realized if pipelines with a small cross sections are chosen, in this case the required power of the pump increases significantly. If no suitable configuration can be found, the distance of the mixing and pumping system to the backfill segment has to be shortened. In this case a pneumatic conveyor system would have to be installed to transport the solids to a mix plant and pumping unit located in the underground galleries to be able to meet the desired backfill rates. In order to identify the optimum solution, the decision criteria and the advantages and disadvantages of existing and planned transport systems were evaluated. The investigations showed that it is technically feasible to pump the backfill material from above ground into the disposal galleries. As failures of the backfill system would result in interruptions of the general disposal operation, a redundant pump and pipeline system was developed. The two lines can be operated separately or simultaneously in order to realize high flow rates needed for the backfilling of segments far from the shaft. (authors)

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DOE OSTI2019

Fundamental Study on Sorption Behavior of Europium to Mica Flakes - 19116

It is well known that biotite exists with flake form in actual underground environment, dominating sorption of cationic radionuclides in a geological disposal system consisting of crystalline rock. The behavior of sorption of radionuclides will differ depending on the difference in shape, while most of the previous studies have used powder biotite. This study focused on the influence of flake structure of phyllosilicate minerals such as biotite on the diffusion and sorption processes of radionuclides, examining the sorption behaviors of Eu on biotite, phlogopite and muscovite in flake form. These flakes (5 mm x 6 mm) were prepared for the sorption experiments and the initial concentrations of the Eu ions were set to 1 mM. The sorption experiments were conducted by mixing 0.4 g of biotite flakes and 4 mL at pH 5 for 7 days. As a result, the sorption amount of Eu increased little by little with internal diffusion in all minerals, respectively. The diffusion process of Eu into the layer structure could be also observed by secondary ion-microprobe mass spectrometry (SIMS) analysis. The apparent diffusion coefficient of Eu into each flake was estimated by fitting a 2-D diffusion model to the sorption data. The estimated diffusion values were clearly smaller than that of matrix diffusion in granite. Hence it was suggested that the so-called Kd-model, which is a local equilibrium model, is not always applicable to the estimation of the retardation effect (i.e., migration rate) of radionuclides in rock matrix. In other words, the migration rate might depend also on the diffusion process of radionuclide in the inside of phyllosilicate minerals in addition of the macro diffusion process in rock matrix. Besides, sorption amount and diffusion coefficient of biotite, phlogopite and muscovite were largely different. Not only the tetrahedral structure adjacent to the potassium but also the octahedral structure might affect the restraint of the interlayer ion (K ions) and the sorption behavior. (authors)

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DOE OSTI2019

Finite Element Structural Analysis Evaluating New Retrieval Strategies in the Hanford Waste Tanks 1 - 19125

Several waste retrieval methods being considered for the Hanford underground single- and double-shell nuclear waste storage tanks would require changes to the equipment loads above the tanks or structural modifications to the tanks themselves. Detailed structural analysis has been used in the planning phase to ensure that the reinforced concrete tanks continue to meet structural design codes in the modified configuration of these waste retrieval options. The feasibility of installing new single and multiple large risers penetrating the tank dome was evaluated using this approach. This paper summarizes the structural analysis methods and design codes used to evaluate the introduction of new unreinforced penetrations in a single-shell tank dome. The anticipated equipment loads during current and future waste retrieval operations are evaluated to ensure the continued structural integrity of the tank under thermal, dead weight, and seismic loads. (authors)

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DOE OSTI2019

Mechanical Waste Gathering System (MWGS), Alternate Retrieval of Hanford Tank Waste - 19188

Hanford has 149 underground single shell tanks; of which several have known or suspected leak sites. Sluicing with large volumes of solution in compromised leaking tanks is not desirable due to the risk of environmental release. To address the complex retrieval of hardpan nuclear waste in Hanford's single shell tanks, a Mechanical Waste Gathering System is being developed and tested, which breaks up the waste and moves it to the surfaces without the introduction of liquid to the tank bottom. To improve the removal of hardpan tank waste, WRPS has leveraged industry knowledge and previous experience to identify a solution. This technology will provide another option for the technology toolbox at Hanford. (authors)

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DOE OSTI2019

Experiments and Simulation of a Borehole in Salt to Understand Heat, Brine, and Vapor Migration - 19192

Disposition of heat-generating nuclear waste (HGNW) remains a continuing technical and sociopolitical challenge. Numerous concepts for HGNW management have been proposed and examined internationally, including an extensive focus on geologic disposal. One proposed geologic material is salt because of its low permeability and viscoplastic deformation that causes self-repair of damage done to the salt by waste emplacement activities. Evaluating the safety and technical challenges of storing HGNW in a salt repository is an ongoing process involving experiments and supporting numerical simulation. Currently an experiment is underway at the Waste Isolation Pilot Plant (WIPP) to explore how the presence of a heat generating source affects phenomena such as brine migration, vapor transport, and mechanical changes to the bedded salt. A sub-horizontal heated borehole test is in progress in the underground at the WIPP that includes a centrally located 10.2 cm diameter borehole with an adjustable heater surrounded with smaller diameter boreholes instrumented with thermocouples. The central borehole contains an inflatable packer, heating block, brine sampler, and constantly flowing nitrogen gas circulation system. Air-injection tests performed in the central borehole provide pressure measurements that are used to constrain permeability of the system. The steady-state temperatures, as well as the rise and fall of temperature when the heater is cycled on and off, have been measured for up to 60 days. In the borehole, dry nitrogen gas circulation evaporates water and outflows to a desiccant container where water mass is measured daily during the experiment to quantify vapor removal. Although this test is a 'shake-down' for a planned second round of fresh borehole testing, we have gathered a rich dataset. These data allow us to build simulations using the Finite Element Heat and Mass transfer code (FEHM) to evaluate the experimental results, determine field-scale parameters, and identify code improvements to reproduce important physical processes that may not be accounted for at present. A 3-D numerical mesh, built using LaGrit software (lagrit.lanl.gov; Miller et al., 2007), includes increasing resolution around the central borehole. Modeling of the experiment allows for determining the local thermal conductivity and permeability of any damaged bedded salt around the borehole, where damage from drilling may change the permeability, porosity and saturation conditions, by parameter testing and inverse methods. Additionally, assumptions about brine and vapor flow and transport are being tested by comparing measured and simulated results. The combination of experimental data and model results provide additional data to help support the safety case for safe and effective HGNW disposition in bedded salt formations. Initial results from this experiment show that water flow into the borehole is within previous experimental results. Further, we have found that the design of the heater block is restricting energy to flow into the rock salt. Thus, this test has proven useful in design of the next generation experiment where infrared heating may be used to bypass issues caused by air gaps located around the current stainless steel block heater. (authors)

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DOE OSTI2019

Concept for Operation of the Belgian Geological Disposal Facility for Category B and C Wastes - 19247

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth of between 200 m and 600 m. Within the framework of the first Safety and Feasibility Case (SFC-1) undertaken by ONDRAF/NIRAS, the organization is updating its former GDF layout, which dates back from 2003. The SFC-1 GDF concept is characterized by a central zone in which shafts, joined by a central connecting gallery, provide two surface connections. On opposing sides of the central gallery, four (two on each side) access galleries branch off to connect the disposal areas for B- and C-waste, respectively. Disposal dead end galleries, 400 m long, branch off perpendicular to the access galleries. Both waste areas will be operated during two separate time frames. The current underground layout has been significantly modified from the previous 2003 layout. The modifications are largely based on several R and D studies with regard to the future operations of the GDF carried out by BGE Technology GmbH in support to ONDRAF/NIRAS. This paper focuses on major operational processes such as waste transport and emplacement. In addition, the current conceptual design of the backfilling and closure concept is presented. Optimization studies on the further development of the operational process are also presented. (authors)

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DOE OSTI2019

Conceptual Design of the Belgian Geological Disposal Facility for Category B and C Wastes - 19341

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF), within poorly indurated clays, as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). The proposed Belgian reference option for disposal of such waste is a GDF within poorly indurated clays, located at a depth of between 200 m and 600 m. Within the framework of the first Safety and Feasibility Case (SFC-1) undertaken by ONDRAF/NIRAS, the former GDF layout, which dates back from 2003 has been updated. The first part of the paper describes the main drivers behind the evolution of the former 2003 GDF design towards the current reference SFC-1 design. It also details how these drivers, related to operational and constructional constraints, have affected the GDF layout, such as the disposal gallery length, the crossing shapes of the GDF galleries and the layout of access and waste shafts. The second part of the paper focuses on the engineering and construction challenges generated by GDF development within poorly indurated clays. This includes discussion of the impact of the specific mechanical characteristics such poorly indurated clays have on the construction methods used for the shafts, the rock support (concrete lining), the drift connections, the crossings and the global GDF design. These construction methods have been selected based on research conducted at the underground research laboratory HADES in Mol (Belgium). Finally, the main axes for further development of the GDF design within the SFC-1 framework is presented. (authors)

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DOE OSTI2019

The Durability of Containers for the Disposal of Radioactive Waste - Swiss and Belgian R and D Studies - 19392

In this paper, we present a summary of the results from research and development (R and D) studies associated with the durability of steel-based disposal canisters for high-level waste (HLW) and spent nuclear fuel (SNF), focusing on experimental studies carried out in support of the Swiss and Belgian disposal programmes. We present the results of laboratory studies carried out for periods of the order of ten to twenty years, and compare them with more recent observations from in situ experiments carried out at the Mont Terri and Grimsel underground laboratories. For the Belgian programme, which is currently envisaging relatively high gamma dose rates outside the steel overpack, we also show the results of studies focusing on gamma radiation at dose rates compatible with those expected in the disposal system. These studies show how, in bentonite systems, corrosion rates decrease to values lower than 1 μm year{sup -1} after an initial period. In the case of cement-based systems, corrosion rates lower than 0.1 μm year{sup -1} are typically observed, with modest effects of gamma irradiation. (authors)

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DOE OSTI2019

Panel Session 32: Panel: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (2/5)

This panel discussed the activities of bringing a first-of-a-kind treatment facility online: The Integrated Waste Treatment Unit (IWTU). The design was designed to treat 900,000 gallons of sodium-bearing liquid waste. Construction of this first-of-a-kind treatment facility was completed in 2012. A new approach was adopted in 2016 after multiple challenges were encountered during facility commissioning. This session focused on the Integrated Waste Treatment Unit (IWTU), which was constructed to convert the remaining 900,000 gallons of liquid sodium-bearing waste at the Idaho National Laboratory Site, Idaho, USA, to a granular solid. The liquid waste was generated during historic spent nuclear fuel reprocessing campaigns at the Idaho Nuclear Technology and Engineering Center and was transferred to three underground storage tanks. The waste was not first-cycle waste but was rather the liquid rinse waste from flushing waste transfer lines between reprocessing runs. IWTU will employ a steam-reforming process to treat the liquid waste. The resulting granular solid will be transferred to stainless steel canisters and then to concrete vaults, which will be stored on site until an off-site national geologic repository is available to accept the waste. Presentation proposed: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (Joel Case, US DOE (USA), Leo Thompson, Sterling Engineering and Consulting Group, LLC (USA), Joseph Giebel, Fluor Idaho, LLC (USA), Richard Boardman, Idaho National Laboratory (USA), Kevin O'Neill, US DOE (USA))

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DOE OSTI2019

Metallic Low-Level Waste (LLW) Handling Options: An Optimal Life cycle-Based Approach for North American Operations - 19413

Daily activities associated with nuclear power plant operations produce large volumes of low-level contaminated waste. During operation, a small percentage of this low-level waste is metallic (< 1% by volume). However, this figure increases materially when a facility is finally decommissioned, with approximately 1000 m{sup 3} of contaminated metal waste generated for every 1 GW decommissioned (13% by volume). With global nuclear decommissioning efforts expected to increase until 2050, the amount of low-level metallic waste generated is expected to increase the global inventory by over an order of magnitude. The high cost of conventional surface storage and burial of radioactive metal, coupled with the projected increase in low-level metallic waste volumes provides strong incentive to identify cost-effective waste management solutions. Surface decontamination and thermal methods are proven options for decontaminating low-level metallic waste. Surface decontamination processes include chemical, electrochemical, or mechanical methods, which are used to remove contamination from exposed layers of metal. Surface decontamination is most effective when metal has low levels of activity localized on the outer layers. When the activity is more uniformly distributed in the metal (or the geometry is complex), alternative methods such as thermal decontamination are required. In the thermal decontamination process, contaminated scrap metal is melted in a furnace and held in a molten state for a period of time while the radioactive material is separated out through gravity. Radioactive material reports to slag which is handled separately, or to the exhaust which is handled using conventional gas handling technologies. Thermal methods of decontamination involving metal melting present an opportunity to reduce long-term storage cost with a reduction in waste volume, along with the opportunity to reuse the decontaminated metal. Methods for dealing with metallic low-level waste from nuclear facilities vary extensively across the globe. The approach employed at any given site may be set by a host of factors ranging from regulatory influence, public perception, local economics, available geology for underground storage and often, precedent established long ago by early operations. This paper examines several low-level metallic waste handling options including techniques for surface decontamination and metal melting, as well as a combination of both. The various methods are categorized according to their suitability for waste characterization and local/global installations. The effectiveness of each option is benchmarked against a range of criteria, including capital and operating costs, environmental impact, overall sustainability and risk management issues. The overall objective of this paper is to develop a methodology for determining the most beneficial low-level metallic waste management strategy for a facility once the quantity, type and nature of the waste is established on a long-term basis. It is the authors' hope that the methodology can be used as a tool to test a facility's current plans (which are primarily focused on long-term storage) and identify areas for optimization with an emphasis on overall cost and environmental footprint reduction. The trend towards long term planning to handle large amounts of contaminated metal, along with obtaining a social license to implement a sustainable waste management solution, are key decision-making variables that will be discussed in relation to the LLW handling options presented in this paper. (authors)

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DOE OSTI2019

Tank Side Cesium Removal System Project and Technology Maturation Program - 19519

Washington River Protection Solutions (WRPS) is the Tank Operating Contractor (TOC) for the U.S. Department of Energy-Office of River Protection (DOE-ORP) on the Hanford Site. The Hanford Site stores an estimated 56 million gallons of mixed radioactive and chemically hazardous waste in large underground tanks. WRPS is in the process of designing the Tank Side Cesium Removal (TSCR) system to produce Low Activity Waste (LAW) from Hanford tank waste. The LAW will be transferred to the Waste Treatment and Immobilization Plant (WTP) LAW Vitrification Facility, where the hazardous constituents in the LAW will be immobilized in a durable glass waste form for disposal. The TSCR Project demonstrates a tank-side treatment system for providing feed to the WTP LAW Vitrification Facility. This system removes cesium from supernatant tank waste using non-elutable cesium ion exchange media to meet the applicable waste acceptance criteria for the WTP. In support of the project, a tailored Technology Maturation/Testing Plan (TMP) has been developed to identify principal Technology Elements (TEs) for the proposed process, designate Critical Technology Elements (CTEs), identify gaps in the existing knowledge base and identify the necessary activities to mature the technologies for successful deployment of TSCR. The project scope and technology maturation approach are presented. (authors)

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DOE OSTI2019

Path to Subsurface Closure: When Real and Predicted Data Don't Match, Central Nevada Test Area (CNTA) - 19560

The Central Nevada Test Area, Nevada, was the site of an underground nuclear test in 1968, which resulted in residual contamination at and near the detonation depth of 975 meters (3200 feet). The environmental corrective action strategy for the site was conducted in accordance with the Federal Facility Agreement and Consent Order and all applicable Nevada Division of Environmental Protection (NDEP) policies and regulations. The corrective action strategy initiated in 1999 included developing a conceptual site model, preparing a numerical flow and transport model, calculating contaminant boundaries, negotiating compliance boundaries with NDEP, performing model validation, and monitoring groundwater. This strategy was executed through the Corrective Action Decision Document/Corrective Action Plan phase, when it was determined that the flow and transport model could not be validated because it failed to adequately predict groundwater elevations in the monitoring/validation (MV) wells (MV-1, MV-2, and MV-3) drilled subsequent to the modeling effort. This led to a change in the corrective action strategy that is now designed to validate the compliance boundary through monitoring and institutional controls, rather than relying predominantly on a numerical flow and transport model. This new approach included enhancements to the monitoring well network and additional monitoring to confirm that data were sufficient to proceed to closure. The corrective action activities were completed in 2015, and the Closure Report, which established the long-term postclosure monitoring program, was approved in 2016. The postclosure monitoring program is designed to (1) assess the effectiveness of the compliance boundary by monitoring for the radionuclides of interest and (2) evaluate the effectiveness of monitoring locations within the groundwater flow system by monitoring water elevations to ensure that monitoring wells are located along the potential migration pathways. (authors)

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DOE OSTI2019

Erosion Challenges at the L-Bar Disposal Site - 19561

The L-Bar, New Mexico, Disposal Site is located in the northeast corner of Cibola County, 10 miles north of Laguna, and was the former site of a uranium mill and underground mine. The mill operated from 1977 to 1981 where a total of 1.9 x 10{sup 6} tonnes (2.1 million tons) of ore was processed. Solid and liquid byproduct, or tailings, from the milling process were impounded behind a dam constructed of native weathered Mancos Shale, which is a predominant geologic formation in the area. The tailings were encapsulated in place, and the resulting 40.5-hectare (100-acre) engineered disposal cell was completed in 2000. The 299-hectare (738-acre) Uranium Mill Tailings Radiation Control Act Title II site was transferred to the US Department of Energy (DOE) Office of Legacy Management under a US Nuclear Regulatory Commission general license in 2004. During reclamation activities, channels were designed and constructed into the surrounding weathered shale to protect the disposal cell from high-precipitation events. Due to the highly erodible soils and anticipated sediment deposition into the diversion channels, a sediment trap was also excavated. By excavating the channels and sediment trap, the local base elevation was altered, which exacerbated erosional forces. Recontouring surrounding drainage basins during the reclamation activities also promoted erosion by reducing natural rock armoring and vegetation that aids in soil stability. To compound the erosion issue, the L-Bar site is susceptible to short-duration, high-intensity, localized convective storms. Because the local geomorphology includes soils consisting of weathered shale, these precipitation events are not conducive to absorption into the soils, but rather to sheet flows that follow topography and concentrate in lower elevation areas where flow volumes and velocities increase. These factors result in the formation of deeply incised gullies and promote soil piping, which is an important geomorphic process in the widening of arroyos and the advancement of head-cutting, evident at the L-Bar site and surrounding areas. At the first annual inspection by DOE in 2005, it was noted that extensive erosion was occurring upgradient of the sediment trap. DOE determined that the design of the site's drainage system was the cause of extensive erosion. It is believed that by lowering the base elevation of the local drainage system through excavation of the sediment trap, the geomorphic response common to fluvial systems in a semiarid environment has increased the incision and remobilization of sediment downstream. Over time, these sediments would be deposited in the channels, which are designed to divert runoff away from the disposal cell, effectively limiting their capacities. The development of alluvial fans on one side of the channel directs flows to the opposite side of the channel, presenting the risk of cutting into the side slopes of the disposal cell. In addition, the upstream migrating gullies could intersect access roads and create hazards for adjacent stakeholders, including their grazing livestock. Current estimates conclude that potential offsite impacts could occur within the next 20 years, with gully migrations of up to 19 meters (63 feet) per year. In 2009, several erosion-control structures were constructed to reduce runoff velocities and ease the elevation differential created by the construction of the sediment trap and diversion channels. The first measurable storm events after construction occurred in August and September 2011, where one structure washed out and another structure was bypassed when runoff breached an earthen berm. DOE has conducted site assessments, continues to monitor erosion, and has applied lessons learned in the process of evaluating options to repair or modify the structures to stabilize the site and prevent potential short-term damage to adjacent properties as well as long-term damage to the disposal cell. (authors)

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DOE OSTI2019

Site Selection Criteria for Deep Borehole Disposal of Intermediate Level Waste in Australia - 19615

Disposal of long-lived intermediate-level waste (ILW) and long-lived spent sealed sources (SSS) requires a high degree of containment and isolation deep underground. Disposal in medium-depth (several hundreds of metres) boreholes in suitable hard rock formations can provide adequate isolation and containment for safe and cost-effective disposal of relatively small volumes of ILW and SSS. In this paper we outline recommended site selection criteria for the purpose of (i) identifying regions that would be excluded from further consideration, and subsequently for the non-excluded regions, (ii) to identify sub-regions that could be developed into a deep borehole disposal site. The CSIRO approach builds on siting guidelines previously developed by the Sandia National Laboratory which have been revised to suit the Australian context. (authors)

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DOE OSTI2020

Alternative Treatment of Defense Waste Processing Facility Recycle via Reuse of Existing Liquid Waste Facilities - 20097

The Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS) has been immobilizing high level waste since 1996. The chemical process within DWPF generates a large volume of condensate, which is recycled to the SRS H-Area Tank Farm. The recycle stream includes a small quantity of sludge solids, as well as soluble cesium that is volatilized during melter operation. The recycle waste is currently received into a large, underground waste tank that separates insoluble solids via decanting. The supernate is treated by an evaporator with a concentrate stream that is stored for future processing. The evaporator overheads are collected and sent to the Effluent Treatment Project (ETP) for final polishing and testing prior to discharge to local surface water. Recycle storage and treatment as described above complicates the overall mission within the tank farms, which are primarily engaged in waste retrieval and preparation activities that support sludge and salt disposition, as well as tank characterization and closure. The need to devote a portion of available storage space to recycle treatment limits operational flexibility, and ultimately the DWPF recycle stream must be diverted to fully close all the SRS waste tanks. An alternative treatment process is being explored to decouple the recycle stream from the tank farm. The proposed treatment process will accomplish solids separation via crossflow filtration and will utilize a wiped film evaporator to volume-reduce the filtrate stream. Evaporator overheads will continue to be further processed in ETP while the solids stream and evaporator concentrate stream will be returned for reprocessing with the DWPF and Salt Waste Processing Facility (SWPF). This process will utilize existing facilities within DWPF and the tank farm that were previously dedicated to Interim Salt Disposition (ISD), but no longer have an identified mission with the startup of SWPF. The reuse of these facilities will remove several constraints from the current Liquid Waste (LW) System Plan without expanding the current footprint of Department of Energy Environmental Management (EM) infrastructure within LW. (authors)

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DOE OSTI2020

Hanford Tank Vapors in Worker Breathing Zones - Source, Dispersion, and Receptor Data - 20373

A wide range of organic and inorganic chemicals from historical Hanford Site processes are now stored in 177 underground storage tanks at the Hanford tank farms. Workers at the Hanford tank farms have expressed concerns about chemical vapor exposures for many years. During the spring of 2014, worker reports of chemical odors and/or symptoms prompted the development of the Savannah River National Laboratory Hanford Tank Vapor Assessment Team (TVAT) and their resulting independent assessment and recommendations concerning tank farm worker reports of vapors. The Tank Vapor Assessment report included a hypothesis that 'vapors coming out of tanks in high concentration (bolus) plumes sporadically intersected with the breathing zones of workers, resulting in brief but intense exposures to some workers.' The focus of this effort is to present current knowledge to describe the mechanisms by which workers may experience short duration vapor concentrations above background in the tank farms environment. The TVAT hypothesis that workers have experienced concentrations approaching 80% of the tank headspace up to 3 meters (10 feet) from tank sources is not supported by sampling data or modeling results. Modeling indicates that concentrations are quickly reduced from the source, and that the upper end of predicted concentrations at worker breathing zones are a factor of 10 or more lower than source concentrations. Area measurements corroborate the fact that worker breathing zone concentrations are 10 to 100 times lower than source concentrations, and events with elevated concentrations are rare. These reduced vapor concentration levels may result in detectable odors or irritation, depending on a worker's specific odor threshold and sensitivity to the chemical species. Short-duration vapor events may be mitigated by evaluating daily atmospheric conditions in conjunction with planned tank farm activities. Monitoring changing conditions related to tank vapors concentrations at the source and within worker breathing zone is also an important step to protect workers from short duration elevated concentration events. (authors)

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DOE OSTI2020

Computationally Aided Design of a Small-Scale Waste-Glass Melter - 20452

To provide mission support to the WTP for the vitrification of legacy nuclear tank waste at the Hanford site, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The waste stored in underground tanks will be separated into HLW and LAW fractions, where the majority is LAW by mass (approximately 90%) and HLW by activity (greater than 95%). The first tank waste to be processed at the WTP will be LAW. Because Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high cost of full-scale radioactive operations and to provide an engineering platform in close proximity to local operations staff. The objectives of this computational modeling effort are to produce process conditions like those encountered in the WTP LAW melters and to minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area or melter-glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models was developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten-glass pool and within the bulk-glass pool. Since melt rate correlates with the velocity underneath the cold cap, this information can be used to select design features that produce the highest gains. (authors)

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DOE OSTI2020

Lessons Learned from Hanford's Purex Tunnel 2 Structural Stabilization Project - 20509

In 2019, U.S. Department of Energy (DOE) contractor CH2M Hill Plateau Remediation Company (CHPRC) completed structurally stabilizing the Plutonium Uranium Extraction Plant (Purex) waste storage Tunnel 2 with engineered grout. This action significantly reduced risk to Hanford workers, the surrounding community and environment by preventing future collapse of the tunnel and possible release of radioactive materials. Decades of plutonium production at the Purex facility required construction of two large underground tunnels for storage of highly contaminated plutonium processing equipment that had failed during the operation of Purex and other facilities across the Hanford Site. Following the partial collapse of Purex Tunnel 1 on May 9, 2017, DOE and CHPRC took immediate action to safely stabilize the collapsed section of the tunnel. The hole in the tunnel was filled the next day as a temporary stabilization measure, and no contamination was detected as a result of the partial collapse of the tunnel. Workers then stabilized Tunnel 1 in its entirety by filling with engineered grout, which also encapsulated the radioactive equipment and railcars within the tunnel. Grouting eliminated the risk of further collapse and stabilized the tunnel contents, all while not precluding future remedial actions or final closure decisions. While conducting an integrity analysis of Tunnel 1 by using construction drawings, photos and other historical documents, project teams also analyzed the much larger Purex Tunnel 2. The analysis found Tunnel 2 did not meet current codes for structural integrity, and it may not be able to bear the weight of the soil above the tunnel. Therefore, it was determined that Tunnel 2 was also at risk of collapse. DoE's independent panel of experts determined grouting to be the best choice for stabilizing Tunnel 2 because it provides the highest level of stability and protection, and does not preclude future remedial actions. DOE authorized CHPRC to commence work and the Purex Tunnel 2 Structural Stabilization Project was launched in December of 2017. The paper will discuss the process of recognizing the following critical aspects and risks associated with stabilizing Tunnel 2, and the actions taken by the project team to mitigate these risks, including: - Tunnel investigation and data collection to assist with future remediation; - On-site batching and mock-ups to provide the large quantity of grout required for stabilizing the tunnel; - Workforce involvement and innovative actions that mitigated the project's most significant potential hazards, including traffic control; - Project team high level of execution translated into successful project execution. Sharing lessons learned from this critical risk-reduction effort can enhance the safety and efficiency of other complex and hazardous projects across the DOE complex. (authors)

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