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DOE OSTI2022

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Advective gas flow in bentonite: Development and comparison of enhanced multi-phase numerical approaches

Understanding the impact of repository gas, generated from degradation of waste and its interaction with the host rock, is essential when assessing the performance and safety function of long-term disposal systems for radioactive waste. Numerical models based on conventional multi-phase flow theory have historically been applied to predict the outcome and impact of gas flow on different repository components. However, they remain unable to describe the full complexity of the physical processes observed in water-saturated experiments (e.g., creation of dilatant pathways) and thus, the development of novel representations for their description is required when assessing fully saturated clay-based systems. This was the primary focus of Task A within the international cooperative project DECOVALEX-2019 (D-2019) and refinement of these approaches is the primary focus of this study (Task B in the current phase of DECOVALEX-2023). This paper summarises development of enhanced numerical representations of key processes and compares the performance of each model against high-quality laboratory test data. Experimental data reveals that gas percolation in water-saturated compacted bentonite is characterised by four key features: (i) a quiescence phase, followed by (ii) the gas breakthrough, which leads to a (iii) peak value, which is then followed by (iv) a negative decay. Three models based on the multiphase flow theory have been developed. These models can provide good initial values and reasonable responses for gas breakthrough (although some of them still predict a too-smooth response). Peak gas pressure values are in general reasonably well captured, although maximum radial stress differences are observed at 48 mm from the base of the sample. Here, numerical peak values of 12.8 MPa are predicted, whereas experimental values are about 11 MPa. These models are also capable of providing a reasonable representation of the negative pressure decay following peak pressure. However, other key specific features (such as the timing of gas breakthrough) still require a better representation. The model simulations and their comparison with experimental data show that these models need to be further improved with respect to model parameter calibration, the numerical representation of spatial heterogeneities in material properties and flow localisation, and the upscaling of the related physical processes and parameters. To further understand gas flow localisation, a new conceptual model has been developed, which shows that discrete channels can possibly be induced through the instability of gas-bentonite interface during gas injection, thus providing a new perspective for modeling gas percolation in low-permeability deformable media.

58 GEOSCIENCES↗
DOE OSTI2022

PFLOTRAN Development FY2022

The Spent Fuel & Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is to develop a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2022 accomplishments by the PFLOTRAN Development group of the SFWST Campaign. The mission of this group is to develop a geologic disposal system modeling capability for nuclear waste that can be used to probabilistically assess the performance of generic disposal concepts. In FY 2022, the PFLOTRAN development team made several advancements to our software infrastructure, code performance, and process modeling capabilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

International collaboration in disposal research: comparative modeling of coupled processes in the DECOVALEX project

Abstract. This presentation gives an overview of an international research collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. The creation of the international DECOVALEX project, now running for more than 25 years, was motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. DECOVALEX emphasizes joint analysis and comparative modeling of state-of-the-art field and laboratory experiments, across a range of host rock options and repository designs. Participating research teams are from radioactive waste management organizations, national research institutes, regulatory agencies, universities, and consulting groups, providing a wide range of perspectives and solutions to these complex problems. The presentation provides examples of the research contributions made collectively in past DECOVALEX tasks and also touches on the unique modeling challenges tackled in the ongoing project phase, referred to as DECOVALEX-2023. The current phase comprises 17 partner organizations, about 50 modeling teams, and 7 modeling tasks, which cover a broad portfolio from fundamental studies on gas migration to full-scale in situ heater experiments in different host rocks to performance assessment studies. Together, these examples illustrate that the insight and scientific knowledge gained within the DECOVALEX project would not have been possible if one group had studied these complex THMC modeling challenges alone rather than within a truly collaborative setting.

Birkholzer, Jens T.↗
DOE OSTI2020

The Influence of Aqueous Dispersivity on Transport of Radiological Material at LANL MDA G - 20450

Aqueous dispersivity is a phenomenon by which a solute moving in groundwater mixes with water that does not contain the solute and becomes diluted. Solutes take paths of varying lengths through pore networks or experience slower velocities due to interactions with pore walls. Aqueous dispersivity leads to breakthrough curves with earlier arrival than expected in purely advective systems and persistent 'tails,' or late-arriving, dilute concentrations of the solute. Dispersivity can be longitudinal (in the direction of flow) or transverse (orthogonal to flow). LANL has disposed of radioactive material at Material Disposal Area (MDA) G since the late 1950's. Most of the disposed materials at MDA G are low-level waste (LLW), but other types of radioactive wastes are disposed or stored at the site. For example, there is a collection of remote-handled transuranic waste (RH-TRU) waste in 33 shafts, which were intended to be stored retrievably. Aqueous-phase radionuclides included in models of contaminant transport at MDA G are subject to dispersive processes in porous media of the vadose zone and saturated zone. To model dispersivity at MDA G probabilistically, we created site-specific distributions of aqueous dispersivity. First, we performed a comprehensive review of peer-reviewed publications, reports from other performance assessments of MDA G, and documentation from sites that were considered similar based on expert opinion, such as Yucca Mountain. Data were included from a wide range of sources, so that distributions developed based on the data reflect the current state of knowledge in properties of the subsurface environment. Data were grouped based on value type (e.g., determined via experiment or used in a different model), rock type, zone type (saturated or unsaturated), and site, and were weighted in the distributions according to their relevance to the site and zone type. We combined Monte Carlo simulation and bootstrapping techniques to produce approximate distributions of mean aqueous dispersivity. Distributions of dispersivity for saturated and unsaturated zone transport are differentiated by data weight and the length scales over which transport occurs. We find that underestimating or disregarding transverse dispersivity may lead to overstating simulated well capture in the saturated zone at MDA G. Neglecting transverse dispersivity is a common assumption due in part to use of one-dimensional transport models that do not consider solute spreading orthogonal to flow. Also, we find that variability in longitudinal dispersivity is associated with substantial travel time alterations in the MDA G unsaturated zone and in simulated contaminant breakthrough times at a downstream observation plane in the MDA G saturated zone. Finally, at large scales (>1000 m), a linear trend between the dispersivity coefficient and length over which transport occurs may overestimate the dispersivity coefficient. Future modeling work with application to solute spreading could include interacting advective and diffusive zones in the highly fractured basalt below MDA G. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Development and Characterization of Cementitious Waste Forms for Immobilization of Granular Activated Carbon, Silver Mordenite, and HEPA Filter Media Solid Secondary Waste

At the Department of Energy’s Hanford site, over 53 million gallons of chemically complex and radioactive wastes have been stored in 177 underground tanks. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is under construction and is designed to treat and immobilize these wastes. During operations of WTP, solid secondary wastes (SSWs) will be generated as a result of waste treatment, vitrification, off-gas management, and supporting process activities. SSW treatment processes and resulting disposal pathways for the final disposition form of the SSW are needed to support direct feed low activity waste (DFLAW) operations and facilitate continued operation of WTP. The SSWs produced through WTP operations are expected to include used process equipment, contaminated tools and instruments, decontamination wastes, high-efficiency particulate air (HEPA) filters, carbon absorption beds (granular activated carbon, GAC), silver mordenite (AgM) and spent ion-exchange resins. These waste streams are planned to be immobilized in a cementitious waste form and disposed of either as stabilized/blended (non-debris) or encapsulated (debris) in a cementitious waste form. Accordingly, cementitious waste forms from these streams were included in the 2017 Integrated Disposal Facility (IDF) Performance Assessment (PA). The input data used to represent these SSW forms in the 2017 IDF PA involved many assumptions and associated uncertainties. This data limitation was due to the lack of material- and site-specific data available for representative SSW materials in cementitious matrices. To verify the assumed values used in the IDF PA and fill this limitation in available data, Washington River Protection Solutions, LLC (WRPS), has initiated a program targeted toward gathering site specific data relevant to Hanford SSW disposal. The work within this report is a continuation of this ongoing program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Biosorption of Uranium by Bone Meal

The treatment of aqueous solution waste containing organic compounds and radionuclides has always been a great concern since the processes available are usually expensive and unfeasible for the low volume of radioactive liquid waste recovered. In view of a more efficient and low cost treatment, biosorption-based technologies for the removal of this hazardous radionuclides, heavy metals, and organic substances from contaminated matrices have been studied as an alternative process. In this context, bone meal is the material of choice, and contains calcium phosphate in the form of hydroxyapatite. Previous studies with hydroxyapatite showed that this material is efficient in removing heavy metals from polluted soils due to its adsorbing ability. In this context, uranium biosorption capacity by using bone meal was evaluated by different experimental conditions. This study aims at analyzing the capacity of bone meal to remove uranium from aqueous solution through biosorption process, considering that this nuclide is radiotoxic and may cause serious environmental problems and risks to human life if not correctly disposed. The biosorption capacity of U by bone meal was evaluated under different experimental conditions. The results suggested that adsorption with this material can be applied in the treatment of aqueous solutions as a low-cost and highly efficient alternative to remove U. Therefore, we believe that our results are useful for other investigators dealing with similar problems. Further work is required to evaluate this material as a potential adsorbent concerning the treatment of liquid organic radioactive waste containing uranium. Next studies are crucial to ascertain the biosorption efficiency, by analyzing other variables, such as adsorbent dosage, pH and temperature.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

DECOVALEX-2023: Task F2-Salt Final Report

The subject of Task F of DECOVALEX-2023 concerns performance assessment modelling of radioactive waste disposal in deep mined repositories. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. In Task F2-(salt), these objectives have been accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks. Coupled-process submodels and deterministic simulations of the entire PA model for a reference scenario for waste disposal in domal salt have been conducted. The task specification has been updated continuously since the initiation of the project to reflect the staged development of the conceptual repository model and performance metrics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

International Collaborations on Radioactive Waste Disposal in Salt (FY20)

This report is a summary of the international collaboration work conducted by Sandia and funded by the US Department of Energy Office (DOE) of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies milestone level-three milestone M3SF-205N010303062. Several stand-alone sections make up this summary report, each completed by the participants. The first two sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS), granular salt reconsolidation (KOMPASS), engineered barriers (RANGERS), and documentation of Features, Events, and Processes (FEPs).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Compaction of crushed salt for safe containment – a summary of the KOMPASS projects

Abstract. For the underground disposal of high-level nuclear waste in rock salt formations, the safety concept includes the backfilling of open cavities with crushed salt. For the prognosis of the sealing function of the backfill for the safe containment of the radioactive waste, it is crucial to have a comprehensive process understanding of the crushed-salt compaction behavior. The crushed-salt compaction process is influenced by internal properties (e.g., grain size, mineralogy, and moisture content) and boundary conditions (e.g., temperature, stress state, and compaction rate) and, therefore, involves several coupled thermal–hydro–mechanical (THM) processes (Hansen et al., 2014; Kröhn et al., 2017). With the paradigm shift from the limited release of radionuclides to safe containment due to the German Repository Site Selection Act passed in 2017, the importance of crushed salt as geotechnical barrier has increased, with a focus on the evolution of its hydraulic properties. Based on the knowledge gaps in the current process understanding, the “Compaction of crushed salt for safe containment” (KOMPASS) projects were initiated to improve the scientific basis behind using crushed salt for the long-term isolation of high-level nuclear waste within rock salt repositories. The efforts to improve the prediction of crushed-salt compaction begun during the first phase of the KOMPASS projects (Czaikowski et al., 2020) and were followed up in a second phase ending in June 2023. The primary achievements of the projects are as follows (Czaikowski et al., 2020; Friedenberg et al., 2022): specification of the KOMPASS reference material, an easily available and reproducible synthetic crushed-salt material, for generic investigations; development of pre-compaction methods and successful production of samples in the short term and under in situ loading conditions; formulation of an extended laboratory program addressing the isolated investigation of known relevant factors influencing the compaction behavior of crushed salt (Düsterloh et al., 2022); execution of long-term compaction tests addressing isotropic and deviatoric load changes, temperature, and compaction state; construction of a backfill body using the KOMPASS reference material in the Sondershausen mine through collaboration with the SAVER (Entwicklung eines salzgrusbasierten Versatzkonzepts unter der Option Rückholbarkeit) project (Schaarschmidt and Friedenberg, 2022); advancement of the tools for microstructure investigation methods (Svensson and Laurich, 2022); generation (first stages) of a microphysical process list combining literature research with our own findings; benchmarking of long-term compaction test for model development and optimization of various existing models as well as the development of new models; application of a virtual demonstrator (2D model representing a backfilled drift in rock salt) for the visualization of developments and the quantification of the models (Rabbel, 2022). In summary, the KOMPASS projects contributed to the reduction of uncertainties and the strengthening of the safety case for using crushed salt within rock salt repositories.

Friedenberg, Larissa↗
DOE OSTI2023

Performance Assessment for the E-Area Low-Level Radioactive Waste Disposal Facility at the Savannah River Site: Appendix H

This section provides supporting material for the development of DU-specific final inventory limits for the GW pathways for every generic waste form and SWF parent radionuclide requiring an inventory limit. The final inventory limits are based on nominal PA transport simulations using PORFLOW as reported in Chapter 5. The nominal PA settings represent the compliance case where some modeling parameter settings are defined based on conservative (pessimistically leaning) arguments. In the overall computational approach employed in this PA, a multitiered radionuclide screening process is adopted as discussed in Chapter 2, Section 2.3.7. In the multitiered process, the initial list of 1,252 parent radionuclides is shortened substantially using conservative, but simple, transport models, along with a reasonably low cutoff criterion of 0.1% SOF value. Multidimensional PORFLOW flow and transport modeling is employed for every parent radionuclide that failed the GW screening. The generic waste form limits represent Tier-3 analyses, while Tier-4 analyses are employed for SWF limits, where warranted.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

New Tank Mapping Method Improves Waste Removal Process

Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.

Mini, Melany↗
DOE OSTI2024

Area V: Waste Management and Subsurface Science (FY21-23 Final Report)

The objective of the Waste Management and Subsurface Science Topic Area V project is to evaluate the feasibility and safety of an intermediate-depth borehole for disposal of Israel Atomic Energy Commission (IAEC) nuclear waste. The IAEC nuclear waste is currently stored in above ground and/or in near-surface facilities that provide short-term radiological worker safety but do not provide long-term, full security and safeguards protection for the nuclear material. Disposal of the IAEC nuclear waste several hundred meters below the surface in a sealed borehole would provide long-term security and safeguards due to its depth underground, as well as long-term radiological safety, due to its isolation in a stable geological formation. This project supports the nuclear security and safeguards initiatives of the Science and Technology Working Group of the NNSA/IAEC Memorandum of Understanding (MOU) and addresses the Topic Area V objective and associated goals. The goals for Topic Area V are: (1) Assess the suitability of the vadose zone in the Yamin Plain of the northern Negev Desert, Israel as a host location for intermediate-depth borehole disposal of radioactive waste, (2) Define and refine the necessary characterization needs to better understand the regional climate, structure, and stratigraphy, geomechanics, geochemistry, hydrogeology, and seismicity of the stratigraphic sequence from the surface to the water table in the vicinity of the proposed host location for the disposal borehole(s), and (3) Develop a preliminary plan and schedule, including necessary resources and costs, for a borehole disposal concept and robust and transparent pre-closure and post-closure safety assessments. The safety assessments will require iteration with process level models at both site and regional scales. Upon completion, the information collected by this effort will allow the IAEC to make decisions regarding the viability and applicability of an intermediate depth (<500m) borehole for disposal in the vadose zone of the Yamin Plain. This FY21-23 final report is provided in lieu of a workshop report that was originally planned for delivery in January, 2024. Due to the conflict in Israel and associated delays, the workshop has been delayed to April 2024. The workshop will be held at LLNL but offsite and will switch to hybrid if travel restrictions remain in place. The associated workshop report will be provided as part of a new FY24-26 LCP.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Development of an ERT‐Based Framework for Bentonite Buffers Monitoring From Laboratory Tests: 2. Quantitative Moisture Dynamics Estimation Model

Abstract The long‐term containment of high‐level radioactive waste in geological disposal repositories relies on Engineered Barrier Systems (EBS), with bentonite clay emerging as a candidate material due to its unique properties. Understanding moisture dynamics within bentonite buffers is crucial for EBS performance, as it directly influences the material's swelling capacity, thermal and hydraulic conductivity, mechanical properties, and long‐term evolution under complex thermal‐hydrological‐mechanical (THM) processes. This study develops an advanced Electrical Resistivity Tomography (ERT)‐based framework to quantitatively monitor moisture dynamics under THM conditions. Our framework extends the Waxman‐Smits model to incorporate the coupled effects of temperature, water content, fluid chemistry, and mechanical changes on bentonite's electrical properties. Utilizing HotBENT‐Lab data from our companion paper, which includes electrical conductivity, CT density, and thermocouple measurements, this study offers a novel methodological framework bridging different scales of the model. Our results show that the extended model can estimate water content from ERT data, capturing spatial and temporal variations in moisture distribution within bentonite columns. However, the model tends to overestimate water content compared to CT density‐derived measurements. We address this discrepancy by incorporating a simplified swelling effect model, which improves agreement between ERT and CT density‐based water content estimates. We also discuss model limitations, including simplified treatment of swelling and micropore effects, and propose a conceptual framework for transitioning from laboratory to field applications, addressing challenges such as parameter scalability, field validation methods, and integration of diverse data sources. This ERT‐based framework can potentially advance real‐world moisture monitoring of bentonite‐based EBS in nuclear waste repositories. Plain Language Summary Safely containing high‐level radioactive waste depends on barriers made from materials like bentonite clay, which is effective because it swells and seals in the waste. To ensure these barriers work well over time, it's important to understand how moisture moves through the clay. Our study developed a new method using ERT to monitor moisture levels in bentonite under conditions that mimic those in actual storage sites, including changes in temperature, water content, and mechanical stress. This study improved an existing model to better account for how these factors affect the clay, allowing us to create more accurate moisture maps. Initially, the proposed model overestimated the amount of water in the clay, but its accuracy was improved by factoring in how the clay swells when wet. This study also identified some limitations of the model and suggested ways to adapt it for use in real‐world waste storage sites. This new approach could lead to better monitoring and safety checks for nuclear waste storage systems, helping to ensure long‐term containment. Key Points This work develops an ERT‐based framework extending the Waxman‐Smits model to monitor bentonite moisture dynamics during coupled THM processes The extended model accurately estimates water content from Electrical Resistivity Tomography data, incorporating swelling effects to improve precision This work proposes a conceptual framework for transitioning from laboratory to field applications, advancing EBS monitoring in nuclear waste repositories

Chen, Hang↗
DOE OSTI2020

Managing Hanford's Direct Feed Low Activity Waste Program - 20002

The Office of River Protection (ORP) is responsible for the management and completion of the River Protection Project (RPP) mission, which comprises both the Hanford Site tank farms operations and the completion and operation of the Waste Treatment and Immobilization Plant Project. The RPP mission is to safely retrieve, treat, and immobilize Hanford's high-level and low-activity tank waste and close the tank farms to protect the Columbia River. The Direct-Feed Low-Activity Waste (DFLAW) Program is a major subset of the overall RPP mission, providing disposition of low-activity waste. The scope is to retrieve, treat, and immobilize the low-activity waste at the Low-Activity Waste Facility. DFLAW will begin operations no later than December 2023. The DFLAW Program overarches a suite of individual projects, activities, and infrastructure upgrades. The major projects and activities include multiple contractors under both ORP and the Richland Operations Office, underscoring the need for effective integration, coordination, and collaboration. The objectives of the DFLAW Program leadership team are to: - Effectively coordinate and integrate the projects that comprise the DFLAW Program - Manage the interfaces between the projects so that the integrated DFLAW Program is completed successfully - Ensure the DFLAW portfolio of projects operate as required without gaps or conflicts at the interfaces. DOE is accomplishing these objectives through a newly established leadership model that integrates the DOE and contractor work streams and promotes a teamwork model to enable collaborative success. Long considered one of the most formidable cleanup challenges at Hanford, the Department of Energy and its contractors are on the verge of achieving a cleanup commitment that has been decades in the making. DOE is preparing to vitrify (turn to glass) Hanford's chemical and radioactive tank waste using the DFLAW process. The Modified Consent Decree milestone for completing hot commissioning of the Hanford Site's Low- Activity Waste Facility is Dec. 31, 2023. While this milestone may seem distant under normal project management circumstances, it is a relatively short period of time for the operational and cultural transformation necessary to successfully begin treating tank waste at Hanford. Direct feed means separating the waste at a tank farm to remove the more radioactive portion (i.e., solids and cesium) so that the resulting low-activity (less radioactive) waste can be fed directly to the Waste Treatment and Immobilization Plant's Low-Activity Waste Facility. Starting DFLAW will require a singular, intensive leadership focus, along with a sustained, collective commitment to excellence and teamwork by the Department and its contractors. Major upgrades to Hanford's infrastructure will occur prior to startup. Supporting DFLAW operations requires significant integration between the DFLAW facilities - the Waste Treatment Plant, Hanford's tank farms, effluent treatment facilities, disposal facilities - and between all of the site contractors to achieve success. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Update on Parallel Process Execution in the Next Generation System Analysis Model

As of the end of 2021, 88,880 metric tons of heavy metal (MTHM) (44,741 MTHM in dry storage; 44,139 MTHM in wet storage) of spent nuclear fuel (SNF) were stored at various reactor sites across the United States [1]. The Office of Storage and Transportation in the Department of Energy is planning for the transportation, storage, and eventual disposal of SNF and high-level radioactive waste (HLW). To aid in this effort and inform decision-makers about the backend of the spent fuel cycle, systems analysis tools capable of analyzing the various options with respect to SNF and HLW management are being used as well as continuously improved to meet the evolving needs of the program. System analysts typically use these tools to vary underlying assumptions (shipping rates, allocation priority, available facilities, start dates, etc.) and study the implications of these changes on site clearance schedules, campaign costs, transportation infrastructure acquisition, etc.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Soil Desiccation Treatability Testing at BC Waste Disposal Cribs

During Hanford’s production period, low-level waste products generated from chemical processing of uranium fuel rods were discharged directly to the ground through a system of cribs and trenches located in the 200-BC-1 Operable Unit (OU). The site consists of 6 cribs and 20 trenches that received more than 117,000 m3 of radioactive liquid waste discharged to the soil. These unlined infiltration galleries held volumes of liquid waste while it seeped into the ground, with the understanding that the 100 m (330 ft) thick vadose zone in the area would effectively capture the effluent and prevent groundwater impacts. A conceptual model showing the operation of cribs and trenches is shown in Figure 1. Data show effluent from the 26 cribs and trenches containing about 410 curies of Technetium-99 (Tc-99) is primarily located between 30 m and 70 m (98 ft and 230 ft) depth (Corbin et al., 2005; Ward et al., 2004). Despite no evidence indicating that the contamination has reached the groundwater at BC cribs and trenches, the mobility of Tc-99 had been demonstrated in laboratory tests and was recognized as a threat to groundwater at the site. Using data from numerical models, laboratory analyses, field investigations, and information on historical discharges, the EPA and Ecology identified Tc-99 and U contamination of the vadose zone as a remediation priority. The U.S. DOE was notified by EPA and Ecology regarding risks associated with Tc-99 contamination in a letter requesting development of a strategy for improved methods to understand the nature and extent of vadose zone contamination, specifically Tc-99, and to develop remedial options for addressing such contamination. To develop the appropriate technology for characterizing, remediating, and monitoring the deep vadose zone Tc-99 contamination, the U.S. DOE worked with the EPA and Ecology to create a Treatability Test Plan under a Remedial Investigation/Feasibility Study (RI/FS) for the Hanford 200 Areas. Under this RI/FS, it was determined that a treatability test for soil desiccation should be carried out as it was identified as a promising in-situ treatment technology for mitigating risks posed by Tc-99 contamination to the groundwater table. The BC Cribs and Trenches site was identified as a representative site for Tc-99 and U contamination and selected for the soil desiccation treatability test. In this chapter, we summarize the overlying regulatory framework of RI/FS and treatability tests and illustrate how development and experimentation supported the evaluation of selected remedies. We briefly discuss the RI/FS for the 200 Areas of the Hanford Site and focus on the soil desiccation treatability testing performed at the BC cribs and trenches site under the Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau (DVZ-TT). The DVZ-TT is one component of the remedial investigation/feasibility study for the Hanford 200 Areas and represents the underlying regulatory framework that drives site operations towards records of decision and site closure.

Mangel, Adam R.↗
DOE OSTI2020

Hydrogen Transport in a Model 9979 Shipping Package with Inner Convenience Cans

Radiolytic hydrogen production and accumulation inside containment packages is a concern at any facility responsible for their packaging, storage, transportation, and/or disposal. When hydrogen gas accumulates to concentrations above the Lower Flammability Limit (LFL) which is 4% or 40,000 ppm in air, the possibility of a deflagration or explosion increases. This concern persists over the course of the package lifetime which is unlimited when disposed of by burial or in permanent repositories. Here, we report on a numerical model used to predict the concentration of hydrogen within each layer of a Model 9979 package containing a convenience can assembly. Simulations show the hydrogen concentration to always be highest in the inner convenience can containing the radioactive source. When the radioactive source is within the Los Alamos National Laboratory (LANL) Packaging Limits, the hydrogen concentration is shown to remain well below the LFL at all times including packaging, storage, transportation, and disposal. A hydrogen transport model is presented for a Model 9979 package system containing a nested arrangement of convenience cans, which are tin oxide coated steel cans of various sizes with a slip-lid assembly. The inner convenience can contains the radioactive source material along with an unknown quantity of incidental water acquired from humid air or processing. While visible organic materials such as paper and plastics were purposely excluded from the inner can, it is not possible to claim the wastes are entirely organic free. The inner convenience can is tape sealed and placed into a plastic bag which is horsetail closed (i.e., twisted and taped). The bagged can is placed into an outer convenience can that is also tape sealed. The can assembly is then placed into the 30 gallon drum and subsequently placed inside the 55 gallon drum in the 9979 package. Here we assume hydrogen gas is produced in the inner convenience can from alpha radiolysis of water at a rate dependent on the quantity of uranium isotopes and water present. The hydrogen transport model was used to calculate hydrogen accumulations within the package’s five layers at different times and conditions. These simulations serve two purposes; (i) to build confidence in the model by comparing predicted values to measured values, and (ii) to check the steady state hydrogen concentrations that are approached at long times in the package’s lifetime. Model simulations were compared to gas samples taken from the 30 gallon drum after storage at LANL’s Chemistry and Metallurgy Research (CMR) building for around 500 days. Hydrogen concentration calculations over much longer periods (i.e., more than 270 years) included extreme storage durations, transportation at extreme cold temperatures, and disposal of packages assuming different average temperatures.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗