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DOE OSTI2016

Prioritizing and Managing Technology Needs to Meet River Protection Project Mission Objectives - 16462

Over 56 million gallons of radioactive waste are stored in 177 large underground tanks at the Hanford Site, located in southeastern Washington State. This nuclear waste resulted from plutonium production for the nation's nuclear defense program. The River Protection Project (RPP) mission is to: - Safeguard and safely manage waste stored in the Hanford tanks, - Treat the waste using the Waste Treatment and Immobilization Plant (WTP), - Ensure safe disposition of the waste to protect the Columbia River and the environment. Each aspect of the mission requires some application of technology in order to meet the mission objectives. The RPP Technology Roadmap document presents a comprehensive, integrated assessment of the technology-related advances needed to ensure successful completion of the Hanford tank waste cleanup mission. This paper describes the development of the RPP Technology Roadmap, summarizes its content, describes integration and collaboration with the Department of Energy's national laboratories to support the work identified, and defines how priorities were identified to ensure near-term RPP mission commitments can be accomplished while preserving the longer term technology development elements necessary for mission completion. The Challenge - Managing a wide variety of technology needs that involve multiple organizations, contractors, and national laboratories presents additional challenges when funding constraints are factored in. A process was needed to ensure the technology needs supporting near term mission priorities were appropriately identified, prioritized, and supported through the fiscal year planning process. Additionally, the national laboratory resources necessary to support many of the technologies require coordination to ensure the most appropriate resources are applied at the appropriate level without duplication or contradiction of effort. The Solution - The RPP Technology Roadmap was developed with the primary purposes to: - Identify technology related items needed to support or enhance the River Protection Project (RPP) mission, - Prioritize technology needs to support fiscal year planning processes, - Organize National laboratory support of the RPP mission and communicate needs to the participating contractors. The Technology Roadmap is consistent with Department of Energy (DOE) Hanford mission priorities and is also in alignment with other Hanford Site technical initiatives and policy and planning documents. The Technology Roadmap highlights the current estimated technology needs and identifies where planning shortfalls exist and where national laboratory participation is needed to support mission success. This information assists fiscal year planning and national laboratory contracting processes to ensure that technology needs are supported. In addition to supporting the RPP baseline mission, the Technology Roadmap provides a basis for strategic planning by identifying opportunities to use technology solutions to enhance mission efficiency. This waste management paper describes the innovative and collaborative processes used to develop, validate, and communicate the results so the project owners and the national laboratories can most effectively plan work scope and resources. A discussion of the following processes is included: - Technology needs and task owner identification, - Functional area grouping and area lead identification, - Needs prioritization, - High priority needs ranking (including validation), - Results compilation and planning gap identification, - Results presentation by project and functional area, - National laboratory coordination and collaboration. Application - The process used to develop the RPP technology roadmap is applicable to any site where multiple national laboratories and contractors are involved in identifying and implementing technology based solutions. This process is also applicable to the DOE HQ role to provide prioritization and coordination of the nationwide technology development activities. (authors)

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DOE OSTI2016

Thermo-Hydraulic-Mechanical Modelling of the Full-Scale Emplacement (FE) Experiment at the Mont Terri Rock Laboratory - 16490

The Full-Scale Emplacement Experiment at the Mont Terri underground research laboratory is a heater test carried out in a gallery constructed in Opalinus Clay. The aim of the experiment is both to develop emplacement procedures for radioactive waste canisters and their surrounding backfill as well as understanding the thermo-hydro-mechanical effects of disposing of radioactive waste on the engineering materials and the host rock. Coupled thermo-hydro-mechanical models have been used to aid planning of the locations of sensors, to inform decisions on the schedule for turning on the heaters as well as to make predictions of the temperatures and water pressures in the bentonite and the Opalinus Clay. The FE Experiment was preceded by a number of other experiments at the Mont Terri laboratory at smaller scales, but with similar materials, which provided crucial input to the models and gave confidence in the general process understanding. Blind predictions were made prior to turning on the heaters, of the temperature and relative humidity in the bentonite surrounding the heaters. The comparison between the blind predictions and early data from the experiment show that knowledge gained from smaller scale experiments on similar materials has enabled good blind predictions to be made and gives confidence that temperature in the experiment is unlikely to rise to a level that could damage the sensors. It also demonstrates that the thermal properties of the disposal system are understood well enough to produce reliable estimates of temperature in the backfill. (authors)

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DOE OSTI2016

Spent Nuclear Fuel Management and Disposal in Finland - Status and Lessons Learned at the End of 2015 - 16507

In November 2015 the Finnish Government approved and the Parliament ratified Posiva's 2012 License Application for Construction of a mined, deep geological repository for the spent nuclear fuel (SNF) at the Olkiluoto site in the Eurajoki municipality based upon recommendations provided by the Finnish Radiation Safety Authority (STUK), international expert groups and long-standing local support. The preceding siting process began in 1983 and it included disposal-concept selection (KBS-3V), 'consent-based' site selection and large-scale in-situ tests conducted in an underground research laboratory, the ONKALO, situated ∼440 m below the mean Sea level immediately adjacent to the planned repository at the Olkiluoto site. At the end of 2015, the Finnish SNF repository is projected to open in 2024 as the first of its kind. Four related lessons learned in Finland during the past 30 years are: 1. The selection of potentially-suitable sites for geological disposal of SNF proved to be more difficult and more time consuming than initially envisioned. 2. The siting of nuclear facilities has evolved from being mainly a radionuclide containment and isolation challenge to becoming a critical moving-target, socio-political challenge governing progress with majority local acceptance being a fundamental imperative. 3. Public and political acceptance and support are strongly linked to the competency, credibility and transparency of the involved organisations and the maturity/pedigree of the proposed disposal concept. 4. Timely sharing lessons learned and conducting selective collaborations with other countries proved to be successful keys to timely and cost-effectively obtain and maintain public and political acceptance and support. (authors)

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DOE OSTI2016

Millennial Time Capsules as a Promising Means for Preserving Records for Future Generations - 16542

When we deal with preservation of Records, Knowledge and Memory (RK and M) for any long-term, non-inspected facility such a deep geological repository, there is no single technical or cultural provision that can be relied upon to do alone the preservation job 100%. Rather we should increase our chances by implementing a combination of approaches based on different components that provide redundancy and/or pointers to one another. Time capsules are no exception. However, they seem well suited to support national archives and other preservation elements, in order to preserve RK and M as long as possible or interesting. Time capsules are a ready to go, workable concept, with many examples of implementation of large-size, millennial time capsules at small depths based on science and lessons to be learned. A proposal is also made for considering and developing small-size time capsules placed strategically deep underground at repository level. (authors)

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DOE OSTI2018

Backfilling of Geological Disposal Facilities - Development of Optimized Backfill Material - 18146

In Belgium, low- and intermediate-level waste (B-waste) will be conditioned in concrete monoliths and high-level waste (C-waste) in so-called Super-containers (SCs). The Belgian radioactive waste management program has focused on the disposal of these waste types in two emplacement fields of a geological disposal facility (GDF). When the disposal waste packages (DWP) have been emplaced, backfilling of the galleries is supposed to be carried out in order to stabilize the galleries, to reduce voids, to guarantee a favourable chemical milieu with regard to the retention of the radionuclides, to minimize a possible gas pressure build-up, to dissipate radioactive decay heat and to enable good conditions for a potential retrieval of the DWP. The current reference solution is to mix the backfill at the surface and then to pump the mixture through pipeline distribution systems into the underground structures. The catalogue of backfill material requirements comprises common, but also different requirements due to the characteristics of the B and C waste packages and differences in the backfilling systems. It was found that two backfill material recipes have to be developed. Different requirements, such as high porosity (low strength and low gas threshold pressure) and high thermal conductivity have contrary consequences for the material. In addition, during backfill production quality variations of the raw materials that influence the backfill material properties can occur. Different lengths of the delivery pipeline may require adjustments of the material composition. In order to consider the practical experiences of backfill material production, one focus of the development programs was to identify compositional ranges that guarantee the respective material property. A combination of the individual 'conformity fields' results in optimal solutions for the standard operating conditions as well as a compositional range that guarantees compliance with the required material specifications. (authors)

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DOE OSTI2018

Determining the Efficiency of the Hanford Waste End Effecter through Proof of Concept Testing - 18165

The ability to effectively retrieve aged and hardened waste from leaking underground storage tanks at the Hanford Site is critical to achieving tank closure and Tank Farm mission success. DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS) is currently developing with Pacific Northwest National Laboratory (PNNL) a confined sluicing end effecter known as the Hanford Waste End Effecter (HWEE). The HWEE is a high-pressure rotating multi water-jet (for waste dislocation) with an eductor transfer system that is designed to both mobilize and extract the eroded waste settled on tank bottoms. This technology development is planned to be accomplished in four phases: 1) HWEE proof of concept testing using cold simulants (based on known Hanford tank waste hardness characteristics); 2) HWEE retrieval system development and testing using cold simulant; 3) full scale HWEE system testing using cold simulant; and 4) deployment into an actual Hanford Tank with the intent to successfully retrieve remaining waste in the tank. This paper discusses the viability of the HWEE based on how well it removed the simulated hardpan and the amount of liquid left in the test bed during Phase 1. Water usage and retrieval rate were closely monitored. This paper only discusses Phase 1 testing and results completed in October 2017. (authors)

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DOE OSTI2018

Sorption of Cationic Radionuclides on Chlorite and Sericite formed by the Hydrothermal Alteration - 18180

Chlorite and sericite formed by the hydrothermal alteration of granite distribute predominantly around the geological repository in deep underground. These minerals often cover the surface of a solid phase and fill the fracture of granite considered as the natural barrier to the migration of radionuclides leaching from radioactive waste. Therefore, to assess the practical performance of the natural barrier, it is necessary to clarify the interaction between radionuclides and these minerals. This study examined the sorption of cationic radionuclides by chlorite and sericite through batch experiments, whereby Cs{sup +}, Sr{sup 2+}, and Eu{sup 3+} were used as tracer ions. In these experiments, powder-form chlorite or sericite was mixed with the solutions of the tracer ions to a liquid/solid ratio of 10 ml/g, and the resulting mixture was allowed to incubate for 2-7 days. The concentration of tracer ions was set to 1 mM and that of NaCl to a specific value in the 0-0.6 M range. The solution pH was 3, 5, or 8. As a result, tracer ions were significantly adsorbed by chlorite and sericite, regardless of pH or NaCl concentration, except when NaCl concentration reached the high value of 0.6 M. The distribution coefficient for europium ions at pH 5 was estimated to be 2-7 mL/g in chlorite and 10{sup 3}-10{sup 4} mL/g in sericite. These values significantly exceeded that measured for feldspar (2-4 mL/g), the predominant component of granite before it undergoes any hydrothermal alteration. These results suggest that cationic radionuclides will be immobilized through sorption in chlorite and sericite filling granite fractures, even if these thermal-fluid-formed minerals cover the surface of granite. (authors)

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DOE OSTI2018

WIPP PA Sensitivity Studies in Support of CRA-2014 Completeness Determination and Compliance Decision - 18209

The Waste Isolation Pilot Plant (WIPP) has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. Containment of TRU waste at the WIPP is regulated by the U.S. Environmental Protection Agency (EPA). The DOE demonstrates compliance with the containment requirements by means of performance assessment (PA) calculations. WIPP PA calculations estimate the probability and consequence of potential radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. Since TRU waste was first received at the WIPP in 1999, compliance recertification applications (CRAs) have been required by the EPA on five-year cycle. The latest CRA (CRA-2014) was submitted by the DOE to the EPA in 2014. As part of the application, a formal set of PA calculations (CRA-2014 PA) was performed and results were documented in detail. The CRA-2014 PA incorporated several parameter updates based on new data, as well as the implementation of a different model for panel closures, barriers planned for the opening of each waste panel. The results presented as part of the CRA-2014 showed via the required complementary cumulative distribution function (CCDF) release curves that the estimated releases were within compliance limits. As part of their process for reviewing the completeness of the CRA-2014 [1], the EPA formally requested that the DOE perform sensitivity studies based on EPA-requested changes to model parameters. The DOE performed and documented the results of three sensitivity studies described below: CRA14-SEN2 [2], CRA14-SEN3 [3], and CRA14-SEN4 [4]. In each case, the new analysis results were compared to CRA- 2014 PA (CRA14) results to determine the impact of the EPA-requested changes. CRA14-SEN2: the application of EPA-requested modified parameters in the operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model resulted in increased pressures and decreased brine saturations in waste areas and increased pressures and brine saturations in the operations and experimental areas. However, the overall effects on total high-probability mean releases from the repository were entirely insignificant. It was concluded that the modeling assumptions associated with the operations and experimental areas of the repository have an insignificant effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN3: the application of EPA-requested modified parameters in the panel closure areas and operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model predominantly result in increased pressures and decreased brine saturations in waste areas, increased pressures and brine saturations in the operations and experimental areas, and decreased brine and gas flows across all panel closures. The pressure increases in repository waste regions yielded slightly decreased brine saturations (on average) in those areas. However, the overall effects on total high-probability mean releases from the repository are entirely insignificant. It was concluded that the modeling assumptions made for this sensitivity study which were associated with the panel closures and north end areas of the repository have only a small effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN4: the application of several changes to the current WIPP PA model (CRA14), including: parameter changes, use of an updated code, and a revised computational grid for the BRAGFLO code. The modifications to the repository model resulted in increased releases in all primary release mechanisms. The impacts of each EPA-requested change to CRA14 were analyzed with respect to each release mechanism in the CRA14-SEN4 study. Overall, total high-probability predicted mean releases from the repository were increased by about 15%. Total low-probability predicted mean releases were increased by about 107%. It was concluded that the EPA-requested changes to the CRA14 result in increases to the predicted total releases from the repository. However, releases calculated in the CRA14-SEN4 analysis remained below regulatory limits, demonstrating continued compliance of the WIPP. In January 2017, the EPA determined that CRA-2014 was complete and then began a technical evaluation of CRA-2014 to determine whether the WIPP continued to comply with the radiation protection standards for disposal. During their evaluation, the EPA utilized the sensitivity study results to inform their ultimate determination in July 2017 that the WIPP continues to comply with regulatory standards. (authors)

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DOE OSTI2018

Nuclear, Land marker for a Waste Isolation Site: International Architecture Competition - 18240

The design and construction of a nuclear waste marker must include many architectural considerations. Very little research has been conducted by architects to address this issue to date. Historically, the task of designating a nuclear waste site has been a discussion predominantly contained within the scientific and engineering communities. In an effort to engage architects within this dialogue, arch out loud hosted the NUCLEAR competition as an international, architectural investigation into the design of a nuclear waste site marker. The NUCLEAR competition asks participants to design a marker to designate the site of the Waste Isolation Pilot Plant (WIPP) located near Carlsbad, New Mexico. The marker will be meant to be a means of passive institutional controls (PICs) to deter inadvertent human intrusion upon the site and should last 10,000 years. The NUCLEAR competition defined the purpose of the nuclear waste site marker as a means to communicate the danger of inadvertent intrusion from future generations. The competition requires that entries must be both legible in communication and durable in their construction. The competition brief challenged participants to consider how customs and languages will evolve over time and to predict what motivations future generations might have for digging at the site. Other background information was provided about the site such as how waste was stored and how the unique geology of the site seals the waste. The 1993 Sandia National Laboratories report titled Expert Judgment on Markers to Deter Inadvertent Human Intrusion into the Waste Isolation Pilot Plant [1] was referenced along with the 1992 WIPP Land Withdrawal Act [2], interviews with WIPP scientists [3] involved with the WIPP PICs program and the Nuclear Energy Agency Records Knowledge and Memory working group. Competition participants were given digital 3-D Computer-Aided Design models for reference that included the underground storage portion of the site, existing buildings and the 16-square-mile site boundary identified in the 1992 WIPP Land Withdrawal Act [2]. This paper will: - Provide a detailed description of the submitted proposals including graphic analysis of the visualizations, diagrams, technical drawings, jury comments and statements. - Catalogue and identify the consistent themes within all submissions with possible categories including: 1) overall architectural form and language, 2) integration of written language and symbols within architecture, 3) use of technology and robotics, 4) construction methods and materials. (authors)

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DOE OSTI2018

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 18327

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to cleanup radioactive waste at the Hanford Site, located in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 56 million gallons of radioactive waste stored in Hanford's 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever, built for radioactive waste processing. DOE is proceeding with its approach to complete the WTP Project in phases, which includes the following principal actions: - Completing construction of the Low-Activity Waste (LAW) Facility, Balance of Facilities (BOF), and Analytical Laboratory (LAB) (collectively referred to as LBL); - Implementing a direct-feed low-activity waste (DFLAW) process to treat low-activity waste independent of the Pretreatment (PT) Facility; - Resolving the technical issues associated with primarily the PT Facility; - Completing the design and construction of the High-Level Waste (HLW) and PT facilities to enable integrated WTP operations. In March 2016, a federal judge amended the 2010 Consent Decree 1 completion milestones for the Hanford WTP, which generally aligns with DOE's plans for completing the WTP facilities in sequence: - Complete LAW Facility hot commissioning: December 31, 2023; - Complete HLW Facility hot commissioning: December 31, 2033; - Complete PT Facility hot commissioning: December 31, 2033; - Achieve initial plant operations for the WTP: December 31, 2036. Having established milestones for completing the WTP facilities has enabled DOE to formally implement its plans for completing the WTP facilities in sequence, starting with the LAW Facility. In December 2016, DOE executed a contract modification and approved a change to the project's performance baseline to complete the LBL facilities and incorporate modifications needed to allow direct feed of low-activity waste to the LAW Facility. Significant progress has been made toward completing the design, construction, and planning for commissioning of the facilities needed to enable the DFLAW operations to commence as early as January 2022. In 2012, DOE partially suspended production work on the WTP's PT Facility and to a lesser degree the HLW Facility because of unresolved technical issues associated with those facilities. In September 2017, DOE acknowledged the WTP contractor had met all the criteria for full resumption of engineering, procurement, and construction (EPC) activities for the HLW Facility. By December 2017, DOE had completed the actions needed to resolve the six of the eight primary technical issues associated with the PT Facility, with the remaining two issues expected to be resolved in the first quarter of 2018. Although significant progress has been made to prepare for resuming production activities for the HLW and PT facilities, DOE has made the decision to constrain EPC work on those facilities in the near term based on DOE's higher priority of completing the WTP facilities needed to begin DFLAW operations. (authors)

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DOE OSTI2016

Computational Fluid Dynamics Modeling of Bubbling in a Viscous Fluid for Validation of Waste Glass Melter Modeling

At the Hanford site, radioactive waste stored in underground tanks is slated for vitrification for final disposal. A comprehensive knowledge of the glass batch melting process will be useful in optimizing the process, which could potentially reduce the cost and duration of this multi-billion dollar cleanup effort. We are developing a high-fidelity heat transfer model of a Joule-heated ceramic lined melter to improve the understanding of the complex, inter-related processes occurring with the melter. The glass conversion rates in the cold cap layer are dependent on promoting efficient heat transfer. In practice, heat transfer is augmented by inserting air bubblers into the molten glass. However, the computational simulations must be validated to provide confidence in the solutions. As part of a larger validation procedure, it is beneficial to split the physics of the melter into smaller systems to validate individually. The substitution of molten glass for a simulant liquid with similar density and viscosity at room temperature provides a way to study mixing through bubbling as an isolated effect without considering the heat transfer dynamics. The simulation results are compared to experimental data obtained by the Vitreous State Laboratory at the Catholic University of America using bubblers placed within a large acrylic tank that is similar in scale to a pilot glass waste melter. Comparisons are made for surface area of the rising air bubbles between experiments and CFD simulations for a variety of air flow rates and bubble injection depths. Also, computed bubble rise velocity is compared to a well-accepted expression for bubble terminal velocity (authors)

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DOE OSTI2018

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

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

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DOE OSTI2018

Basis of Knowledge for Oxidizing Chemicals in TRU Waste - 18654

A new Waste Isolation Pilot Plant (WIPP) document for implementing new requirements in the WIPP Waste Acceptance Criteria (WAC) titled Basis of Knowledge for Evaluating Oxidizing Chemicals in the TRU waste was issued in July, 2017 and has now been implemented on several waste streams. The February 14, 2014, airborne radiation release in Room 7 of the WIPP underground disposal unit 7 was caused by an exothermic chemical reaction of nitrate oxidizing chemicals and organic materials in a single 55-gallon container. That event exposed hazards not previously evaluated in the WIPP safety basis. A new WIPP Documented Safety Analysis (DSA), Revision 5b, was developed that addressed potential hazards related to the airborne radiation release and evaluated and addressed new fire hazard scenarios. The revised hazards analysis within the DSA specifically included potential chemical exothermic reactions in waste containers and propagating fires. Compliance with WAC was credited as initial conditions to the hazards analysis, and Chapter 18 was added making the WIPP WAC compliance program a safety management program within the DSA. The WIPP DSA, Revision 5b, Chapter 18, and the WIPP WAC require the use of this document when performing the enhanced acceptable knowledge (AK) process on transuranic (TRU) waste determined to contain oxidizing chemicals. The WIPP WAC was revised to address the conditions and requirements of DSA Revision 5b. The use of this Basis of Knowledge (BoK) by WIPP Certified Programs is specified in the WIPP WAC, Appendix H, Enhanced Acceptable Knowledge. This document establishes criteria that must be used by the WIPP Certified Programs to evaluate TRU waste containing one or more oxidizing chemicals to determine acceptability at the WIPP as-is and to identify when additional evaluation or treatment is required. This BoK also includes options and requirements to be used by TRU waste sites when further evaluation, testing, and/or treatment are required. The U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) will review requests, plans, and technical justifications and provide written decisions prior to implementation by the TRU waste sites. (authors)

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DOE OSTI2018

Enhanced Evaluation Criteria to Meet the WIPP Waste Acceptance Criteria - 18655

The Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the disposal of Transuranic (TRU) waste and the only such disposal facility of its kind. TRU waste is radioactive mixed waste generated by the U. S. Department of Energy in support of National Defense. In 2014, the WIPP facility was shut down when two separate accidents occurred in the repository. The two accidents were a fire on a large underground haul truck that prompted evacuation and filled the mine with smoke and soot, and a radiation release that contaminated the exhaust ventilation and disposal areas of the mine. DOE performed two accident investigations in 2014 that resulted in many corrective actions. Some of the corrective actions resulted in significant changes in the WIPP Waste Acceptance Criteria (WAC) and are required to be applied to all TRU waste upon the resumption of shipments to WIPP. WIPP re-opened and began taking shipments again in 2017 and a large backlog of over 25,000 TRU waste containers that were certified prior to the new WAC have been undergoing re-evaluation since the re-opening. This paper will focus on the new WAC criteria that TRU waste must meet above-and-beyond previous WAC criteria and the implementation of new processes for approval. The primary area for these changes is in the area of waste characterization using Acceptable Knowledge (AK) and new detailed analyses required for chemical compatibility. The accident investigation revealed that the radiation release was due to a thermal runaway and bursting of the waste container due to inadvertent mixtures of nitrates and combustibles resulting in ignitable waste forms. In other words, the TRU waste container had incompatible contents that went undiscovered and created a hazard characteristic that was prohibited at WIPP. Through detailed chemical compatibility evaluations and new oversight, the DOE should be able to prevent a similar accident (see Figure 1). The new chemical compatibility evaluations and evaluations of oxidizing chemicals combined with combustibles are complex and detailed. Chemical compatibility is not a new requirement, but a much more detailed evaluation is a new requirement. And, obtaining historical information about the generation of the final waste form for legacy packaged waste can be very challenging. Specifically, determining possible amounts of chemicals in the waste can be extremely challenging. This paper does not explore in depth, the details of oxidizing chemicals. Other papers presented will discuss oxidizing chemicals. (author)

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DOE OSTI2018

Meeting the Challenges, Exceeding Expectations in Radioactive Liquid Waste Work - 18657

The Savannah River Site (SRS) is an 803-square kilometer (310-square mile) U.S. Department of Energy (DOE) site located near Aiken, South Carolina, encompassing parts of Aiken, Barnwell, and Allendale counties. SRS is a key industrial complex responsible for environmental stewardship, environmental cleanup, waste management, and disposition of nuclear materials. The Site was constructed in the early 1950's in support of the United States' Cold War efforts. In July 2009, the DOE began a new era by awarding a stand-alone contract to execute the Liquid Waste Mission at SRS. The Liquid Waste Contractor is responsible for all liquid waste operations, including waste management, disposition, and management of all liquid waste facilities at SRS, as well as operationally closing high-level waste tanks. Currently, there are 43 operational underground storage tanks containing approximately 132.5 million liters (35 million gallons) of radioactive waste. Before tanks can be closed, work involves removing, stabilizing, and disposing of high- and low-level radioactive waste from the tanks. (authors)

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DOE OSTI2018

River Protection Project System Plan, Rev. 8 - 18661

The River Protection Project System Plan comprises a set of scenarios that evaluate options for safely treating the high-level radioactive waste stored in underground tanks at the DOE Hanford Site. It establishes a baseline case, used for communicating the DOE's strategy to various stakeholders, as well as ten alternative scenarios that evaluate initiatives to potentially accelerate cleanup and/or address risks. The scenarios were modeled using TOPSim, a custom flowsheet simulator that includes chemistry and mass transfer elements used to estimate operating durations and completion dates, quantities of waste treated through various processes, as well as the time required to complete treatment. The TOPSim results are aligned to a resource-loaded Oracle{sup R} Primavera P6 schedule to estimate the life cycle cost of the cleanup mission. This paper will outline the results of the System Plan, Rev. 8 modeling effort. (authors)

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DOE OSTI2019

Development of a Trace Detection Method for Mercury Speciation in Hanford Tank Waste - 19014

The U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State stores 56 million gallons of chemical and radioactive waste stored in underground tanks. Mercury species in liquid and solid fractions of Hanford tank waste pose a risk to the implementation and execution of DoE's River Protection Project mission due to mercury concentration limits imposed under waste acceptance criteria, toxicity characteristic leaching procedure restrictions for secondary waste disposal, and air emission permits. Although there is currently no processing of tank waste at Hanford, waste processing is imminent, and representative mercury inventory tracking will be important to anticipate and mitigate stack emissions and secondary liquid waste disposal issues. A methodology to accurately determine total mercury and mercury speciation in Hanford tank waste, while maintaining low mercury detection limits, currently does not exist at Hanford. Moreover, the mercury species in the waste solids are unknown and may be waste type-specific. The development of an analytical method to separate and speciate mercury in tank waste was undertaken by the Pacific Northwest National Laboratory (PNNL) while leveraging an existing non-radiological, ultra-trace mercury detection technique frequently performed at PNNL's Marine Sciences Laboratory in Sequim, Washington. The ultimate goal of this work is to develop an analytical method to obtain detailed mercury speciation information and update flowsheet modeling assumptions related to mercury partitioning at Hanford. The first phase of testing was completed in the development of an analytical methodology for the low-level determination of total mercury, elemental mercury, methylmercury, and dimethylmercury in Hanford tank waste. This paper presents Phase I testing which focused on determining whether mercury species can be preconcentrated from aqueous Hanford tank waste solution onto a solid substrate, while simultaneously eliminating or very significantly reducing the parent radionuclide activity. This was demonstrated by developing protocols for the separation and preconcentration of total mercury and methylmercury onto solid substrates, then measuring the substrates for radiological content at PNNL's Radiochemical Processing Laboratory on the Hanford Site. The mercury content was successfully separated onto stable substrates and the radionuclide activity reduction was significant enough that the preconcentrated samples could be subjected to quantification using far less restrictive handling and radiological controls than would be required for the parent sample (i.e., contamination area fume hood vs. hot cell). This mercury separation method required no sample dilution, which enables maximum mercury detectability, while still allowing mercury speciation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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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