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DOE OSTI2017

Special Analysis: 2017-001 Disposal of Drums Containing Enriched Uranium in Pit 38 at Technical Area 54, Area G

Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. This special analysis, SA 2017-001, evaluates the potential impacts of disposing of this waste in Pit 38 at Area G based on the assumptions that form the basis of the Area G PA/CA. Section 2 describes the methods used to conduct the analysis; the results of the evaluation are provided in Section 3; and conclusions and recommendations are provided in Section 4.

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

Radionuclide Inventory Threshold for Trench 31 and 34 Located Within the 200 West Area Low-Level Burial Grounds

Trenches 31 and 34 located in the 200 West Area Low-Level Burial Ground (LLBG) are authorized for disposal of low-level and mixed low-level radioactive waste under DOE Order 435.1, Radioactive Waste Management. A revised Disposal Authorization Statement was issued in October 2023 following approval of the performance assessment (PA) and supporting technical basis documents. To be consistent with approved PA calculations, the waste acceptance criteria (WAC) needs to be updated to include concentration limits for Category 1 (CAT1) and Category 3 (CAT3) waste along with the inventory limits. This environmental calculation file (ECF) provides the inventory limits for Category 1 and Category 3 waste disposal of radionuclides at Trench 31 and 34 based on the PA calculations. The concentration limits are not calculated as they are provided in the PA itself. The inventory thresholds (or limits) are calculated for only those radionuclides that are identified in the PA as the primary dose contributors. The primary dose contributors for the groundwater pathway are technetium-99, iodine-129, niobium-94, chlorine-36, molybdenum-93, and niobium-93m while those for the air pathway are tritium and carbon-14.

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DOE OSTI2018

The Retrieval project Asse II - Lessons Learned in Radiological Protection - 18263

From 1967 to 1978 in total 125,787 drums and packages with low- and intermediate level waste were emplaced in the Asse II mine. The disposal operation ended in 1978 due to a revision of the Atomic Energy Act. For the disposal of radioactive waste, a plan approval procedure would have been mandatory due to the Atomic Energy Act, but the mining law was the legal basis for operating the Asse II mine, further on. A concept for a closure of the mine was not existing at that time. On January 1, 2009, the Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz BfS) became the new license-holder and operator of the Asse II mine under the Atomic Energy Act. The Act on the Organizational Restructuring in the Field of Disposal in Germany became effective on July 30, 2016. The act provides that a federally owned company, the Bundesgesellschaft fuer Endlagerung (BGE mbH, Federal Company for Radioactive Waste Disposal) is formed to carry out the operative tasks of searching for a High Level Waste (HLW) repository, the construction and operation of repositories (Konrad, Morsleben) and of the Asse II mine. Since April 25, 2017 BGE is in charge of the operative tasks that have so far been carried out by BfS. Up to now, the long-term safety of the Asse II mine has not been demonstrated in the case that all the waste will remain inside the mine. Further problems arise from a lack of host rock stability, the incomplete backfilling of the mining areas that have not been used for the emplacement of radioactive waste, and a continuous brine inflow into the mine of approx.12 m{sup 3} per day. For these reasons, according to chap. 57b of the Atomic Energy Act, the Asse II mine is to be closed without delay after retrieving the emplaced nuclear waste. Besides the recent organizational changes in Germany and the special problems giving rise to retrieve the waste, in this paper the operational experiences since 2009 on the way from mining law to atomic law will be described. This applies to the radiation protection regime, the release and clearance procedure, and some measures to boost efficiency. (authors)

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DOE OSTI2019

Developments in the Regulatory Framework of Nuclear Waste Management in Germany - Licensing Procedures for storage of SNF and Public Involvement - 19612

According to article 6 of the German Atomic Energy Act (AtG), storage of Spent Nuclear Fuel requires a license. The competent licensing authority in this field is the Federal Office for Safety of Nuclear Waste Management (BfE). According to the Act for selecting a repository site for heat-generating radioactive waste (HAW) (Site Selection Act - StandAG) dating from 2017, BfE has been assigned the task of developing, conducting, and evaluating public participation procedures in the process of site search and selection for a deep geological repository for radioactive waste and SNF. Article 6 AtG applies to interim storage of SNF. If the legally defined licensing conditions by article 6 (2) AtG are met, the storage license applied for has to be granted. For on-site facilities, interim storage has been limited to 40 years. The licensing procedures for interim storage of SNF are carried out in accordance with the general administrative procedural rules. In certain cases due to the Environmental Impact Assessment Act (EIA Act), an EIA has to be carried out. In these cases, a formalized licensing procedure including an involvement of the public is conducted. The Nuclear Licensing Procedure Ordinance (AtVfV) originating from 1977 and reformulated in 1995 defines formalized rules for the public participation procedure which have to be followed strictly. These include a public hearing. The risen objections are evaluated by the licensing authority and taken account of in the licensing decision. The process of repository site selection defined by the StandAG exceeds these formalized procedures significantly: The objective of this process is to achieve a broad social consensus about the site selection procedure and about a repository site that will offer the best possible safety for one million years. Therefore, the process has to be particularly transparent, and involvement of the public - including newly developed forms of public involvement - is an essential part of it. BfE as competent authority in both the fields of repository site search and selection and licensing of interim storage of SNF has recently developed new strategies of involving the public concerning interim storage of SNF. Supplementing the standard licensing procedures, BfE presently informs and involves the public about topical issues regarding the present state and recent developments in the field of interim storage of SNF. Current central topics are for example the return of HAW and MAW from reprocessing of SNF in France and UK as well as perspective longer interim storage periods on the interface with final disposal. Regarding longer interim storage, host communities of interim storage facilities are concerned that interim storage might be extended indefinitely. Additionally, there is a strong fear of transports of radioactive waste and SNF from one place to another. (authors)

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DOE OSTI2015

Field Test to Evaluate Deep Borehole Disposal.

The U.S. Department of Energy (DOE) has embarked on the Deep Borehole Field Test (DBFT), which will investigate whether conditions suitable for disposal of radioactive waste can be found at a depth of up to 5 km in the earth’s crust. As planned, the DBFT will demonstrate drilling and construction of two boreholes, one for initial scientific characterization, and the other at a larger diameter such as could be appropriate for waste disposal (the DBFT will not involve radioactive waste). A wide range of geoscience activities is planned for the Characterization Borehole, and an engineering demonstration of test package emplacement and retrieval is planned for the larger Field Test Borehole. Characterization activities will focus on measurements and samples that are important for evaluating the long-term isolation capability of the Deep Borehole Disposal (DBD) concept. Engineering demonstration activities will focus on providing data to evaluate the concept’s operational safety and practicality. Procurement of a scientifically acceptable DBFT site and a site management contractor is now underway. The concept of deep borehole disposal (DBD) for radioactive wastes is not new. It was considered by the National Academy of Science (NAS 1957) for liquid waste, studied in the 1980’s in the U.S. (Woodward–Clyde 1983), and has been evaluated by European waste disposal R&D programs in the past few decades (for example, Grundfelt and Crawford 2014; Grundfelt 2010). Deep injection of wastewater including hazardous wastes is ongoing in the U.S. and regulated by the Environmental Protection Agency (EPA 2001). The DBFT is being conducted with a view to use the DBD concept for future disposal of smaller-quantity, DOE-managed wastes from nuclear weapons production (i.e., Cs/Sr capsules and granular solid wastes). However, the concept may also have broader applicability for nations that have a need to dispose of limited amounts of spent fuel from nuclear power reactors. For such nations the cost for disposing of volumetrically limited waste streams could be lower than mined geologic repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Qualification of Numerical and Semi-Analytical Modeling Codes for Thermal Design of HLW Repository - 18455

Disposal in a geological repository is a viable option for safely isolating high-level radioactive waste. Thermal processes induced by decay heat from the radioactive waste may impact the integrity of the engineered barrier system, and may lead to fluid flow and transport; and the associated migration of radionuclides. This study looks at an analysis of thermal conduction for the emplacement of high-level waste in a repository located in bedded salt. The simulations used analytical and numerical methods to study conductive heat transport in a repository. This study concentrated on heat flow in a central waste package with contributions from adjacent waste packages and drifts. Of particular importance to this work is comparison of predictions of peak temperature at the drift wall and at the waste package surface using different solution methods. Simulation results of analytical and numerical modeling methods are compared. The results show that predictions of the codes were comparable under identical initial and boundary conditions. (authors)

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DOE OSTI2019

Low-Level Waste Minimization and Management Planning at the Western Waste Management Facility, Ontario Power Generation - 19235

The purpose of this study was to examine the strategies intended to mitigate the temporary storage restrictions for Low-Level radioactive waste at the Western Waste Management Facility to enable continued accommodation of waste generation rates, and to recover storage capacity from 2018-2023 to optimize long-term sustainability for radioactive waste management. The management options considered were based on existing waste processing technologies and programs available at the Western Waste Management Facility, in combination with third party volume-reduction services. Completed analysis measured the consumption of Low-Level Waste storage space as influenced by the projected volumes from nuclear generating stations, and the various processing scenarios developed; such as reducing the volume of surplus processable waste on-site, and the reduction of specific waste streams previously considered to be non-processable. Results from the quantitative analysis indicated which strategies will be most effective for minimizing waste volumes for long-term storage and for recovering low-level storage space at the Western Waste Management Facility. The strategic path forward is presented in short- to long-term management objectives. Results from this study will contribute to the waste stream planning and management of major reactor refurbishment projects and to the technological advancements and innovation of radioactive waste processing and minimization. With further understanding of the refurbishment-specific waste types and volumes, it is expected that Nuclear Waste Management and the nuclear generating stations will achieve increasingly improved minimization results with the progression of each refurbished reactor. (author)

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DOE OSTI2017

Problems Related to Temperature-Induced Processes in Long-Term Repositories for Nuclear Waste - 17514

Concepts for the final disposal of high level radioactive waste in deep geological formations must consider the heat pulse produced from radioactive decay during operations and in the post-closure phase as an important factor for the integrity of barrier systems. Several safety cases in different countries limit the maximum temperature at the boundary between the technical barrier/canister surface to the host rock/buffer material to 100 C. Increases in maximum temperature have very diverse effects with respect to the type, intensity and range of the heat pulse. These depend, among other factors, on i) the amount of heat emitted, ii) the type of host rock including site-specific characteristics, iii) the safety and repository concept and further site-specific boundary conditions. The German Commission on the Disposal of High-level Radioactive Waste (Commission) recommended, for precautionary reasons, that the temperature limit at the outer surface of waste containers should not exceed 100 C. Higher temperatures could only be considered if the maximum physically possible temperatures in the respective host rock have been reliably set as a result of research findings. (authors)

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DOE OSTI2020

Utilizing Commercial Submersible Mixer Pumps for Sludge Removal in Savannah River Site's Tank 26 - 20289

The Savannah River Site (SRS) Liquid Waste System (LWS) safely manages, stores, treats, and dispositions liquid radioactive waste. The LWS consists of 51 underground waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities, known as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). One of the waste storage tanks, Tank 26, was placed into service in 1980 as an F Tank Farm (FTF) Evaporator Feed Tank. From 1980 to 2013, Tank 26 received F Canyon receipts and dilute supernate that was then transferred to the 242-16F FTF Evaporator for volume reduction. In 2013, the steam tube bundle failed in the FTF Evaporator, and the decision was made not to repair/resume evaporator operation. During operation of Tank 26, solids built up to a level of 2.03 m in the tank with a volume of 1062.9 kL. Due to the evaporator failure and subsequent shutdown, the solids in Tank 26 were selected to feed Sludge Batch 10, which is collected and prepped prior to being sent to DWPF for final disposition. To accomplish solids removal, Commercial Submersible Mixer Pumps (CSMPs) were selected to slurry the solids in preparation to be sent to Tank 51. Four CSMPs were installed in Tank 26 with each one installed in a separate quadrant of Tank 26. This work would mark the first deployment of CSMPs in an SRS waste tank. The CSMPs were developed in response to operational issues from previous mixing pumps and budgetary constraints. The CSMPs use the concept of modifying commercially available equipment for nuclear waste applications. The CSMPs consist of a 230-horsepower submersible mixing pump (manufactured by GPM, Inc.) joined to a mast fabricated at SRS. For this application, Savannah River Remediation (SRR) design services was challenged to provide a mast design that required no steel superstructure support system on top of Tank 26. Also, SRR design services provided a simplified Tank 26 riser interface that did not require extensive machine parts to adjust height of the suction screen of the CSMP within the tank. Other design improvements include no requirements for tank top radiation monitors and no requirements for flushing of the CSMPs during startup/shutdown operations. The CSMPs were operated at an initial height of 2.29 m above the Tank 26 bottom for 10 days. After sludge sounding, the CSMPs were lowered to a height of 1.52 m with no issues due to proper work planning and the simplified riser interface. With the CSMPs lowered to a level of 1.52 meters above the tank bottom, the CSMPs were operated for 12 days and another sounding was performed. The sounding level came back matching the disturbance depth results of the first pump run. Chemistry samples were analyzed while the tank was left to settle. The analysis revealed enough weight percent solids to provide good feed to Sludge Batch 10. Also, the results revealed the supernate within the mixture would provide good salt solution feed to Salt Waste Processing Facility (SWPF) Salt Batch 3. So, the plan was modified to let the disturbed solids settle and decant the salt solution for addition to SWPF Salt Batch 3. Then water will be added back to Tank 26, and the CSMPs run in order to wash the solids to decrease settling time prior to being sent to Tank 51 for Sludge Batch 10. Utilizing CSMPs for waste removal in an SRS Tank has provided a cost-effective means for further waste removal efforts. In addition, the CSMPs are easier to operate by utilizing a robust and simplified design. The CSMPs performed quite well with no process shutdown or delays during operation. As a result, CSMPs are integral to the future of removing radioactive waste from storage tanks at SRS. (authors)

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DOE OSTI2026

Annual Summary Report for the Remote-Handled Low-Level Waste Disposal Facility—FY 2025

The U.S. Department of Energy (DOE) requires the performance assessment (PA) (Department of Energy Idaho Operations Office [DOE-ID] 2018a), composite analysis (CA) (DOE-ID 2012), and CA addendum (DOE-ID 2018b) for the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility at the Idaho National Laboratory (INL) Site shall be maintained to evaluate changes that could affect the performance, design, and operating basis for the facility (DOE Manual 435.1-1 Change 3, “Radioactive Waste Management Manual,” Section IV.P. [4]). The RHLLW Disposal Facility became operational in September 2018 after the completion of operational readiness activities required by DOE Order 425.1D, “Verification of Readiness to Start Up or Restart Nuclear Facilities,” and the issuance of the startup authorization by the Startup Approval Authority (Boston 2018). The first waste disposals at the RHLLW Disposal Facility began in Fiscal Year (FY) 2019. In Fiscal Year (FY) 2025, no significant operational changes or other activities occurred that would cause deviation from the assumptions in the PA and CA pertaining to disposal geometry, verification of waste characteristics, tracking disposal inventories against total limits, facility closure design, or institutional controls. This FY 2025 annual summary report (ASR) documents the continued adequacy of the PA, CA, operating disposal authorization statement (ODAS) (ODAS 2018), ODAS technical-basis documents, and the radioactive waste management basis (RWMB) (INL 2024a) to meet DOE Order 435.1, “Radioactive Waste Management,” performance objectives for the RHLLW Disposal Facility. Annual review of the adequacy of the PA and CA at the RHLLW Disposal Facility ensures that conclusions of the analyses remain valid in accordance with requirements of DOE Order 435.1.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2021

Forty years of durability assessment of nuclear waste glass by standard methods

Abstract Standard methods to assess the durability of vitrified radioactive waste were first developed in the 1980’s and, over the last 40 years, have evolved to yield a range of responses depending on experimental conditions and glass composition. Mechanistic understanding of glass dissolution has progressed in parallel, enhancing our interpretation of the data acquired. With the implementation of subsurface disposal for vitrified radioactive waste drawing closer, it is timely to review the available standard methodologies and reflect upon their relative advantages, limitations, and how the data obtained can be interpreted to support the post-closure safety case for radioactive waste disposal.

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DOE OSTI2019

Safety Classification and Design Rules for Structures, Systems and Components in a Near Surface Disposal Facility-Post Closure Phase - 19716

Canadian Nuclear Laboratories (CNL) is developing a Near Surface Disposal Facility (NSDF) to be constructed at its Chalk River Laboratories (CRL) site. In general, a NSDF is a facility for radioactive waste disposal, under an engineered cover, located at or within a few meters of the earth's surface. The NSDF will establish a safe, local and permanent means for the disposal of low level radioactive waste. The NSDF is intended to dispose historic waste currently in interim storage as well as waste arising from building decommissioning and environmental remediation activities. The NSDF is to be composed of engineered containment mound, waste water treatment plant and other supporting facilities and infrastructures that are needed to preserve radioactive waste. The NSDF life cycle will comprise two distinct phases: pre closure and post closure. The pre closure phase involves construction, operation and closure activities. The post closure phase includes institutional control and post-institutional control to ensure long term safety and radiological acceptability of the disposal facility. As part of the evolving activities for the NSDF project, a process is developed in accordance with Canadian regulations and International Atomic Energy Agency (IAEA) principles to classify systems that comprise NSDF for safe operation, long term safety of the disposal facility and the protection of workers, the public and the environment. This paper describes the methodology developed to classify Structures, Systems and Components (SSCs) important to safety in post closure phase and determine their design rules. The design rules are the relevant codes, standards and proven engineering practices that should be applied, as appropriate, to the design of SSCs to meet the applicable design requirements. (authors)

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DOE OSTI2019

Developing Waste Characterization Capabilities in Support of the Near Surface Disposal Facility - 19245

Upon approval from the Canadian Nuclear Safety Commission, Canadian Nuclear Laboratories (CNL) Proposed Near Surface Disposal Facility (NSDF) to be constructed at CNL's Chalk River Laboratories will be the first permanent disposal facility in Canada for the disposal of low level radioactive waste. The NSDF is currently in the approval process and as such, a preliminary Waste Acceptance Criteria (WAC) has been implemented for the proposed facility. Waste currently generated at CNL is being characterized to verify that it meets the WAC and will be 'disposal ready' once the NSDF is operational. A robust waste characterization program is required for the following reasons: 1. Improved waste segregation to support waste minimization and recycling to ensure non-radioactive waste is reliably segregated and removed from radioactive waste streams destined for the NSDF. This includes characterization and discrimination of Naturally Occurring Radioactive Material (NORM) and anthropogenic radioactive material. 2. To ensure low level waste (LLW) is properly characterized as per the limits stipulated within the WAC to ensure criteria is met for NSDF packaging requirements 3. To ensure intermediate level waste is reliably segregated from NSDF destined wastes streams The purpose of this paper is to provide a review of the development of a waste characterization program strategy at Chalk River Laboratories and applicability of various waste streams; including consideration of important factors such as measurement methods, techniques, and the management of data. Emphasis is necessarily placed on non-destructive gamma spectrometry systems which are central to CNL's approach. A review of near-term developments is also provided. (authors)

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

Algebraic expressions for estimating the impact depths of a surface barrier over a homogeneous soil

Engineered surface barriers are used to isolate subsurface contaminants for effective long-term containment of municipal solid waste, other nonhazardous solid and liquid waste, hazardous and toxic wastes, and radioactive waste. The impact depths of a surface barrier are affected by the pre-barrier recharge rate and the properties of the soil beneath the barrier. In this paper, the pore-size-specific (PSS) water velocity is defined and an algebraic expression of PSS velocity is derived based on the stream tube concept and the Brooks and Corey hydraulic retention model. Algebraic expressions are developed to estimate drainage velocities and barrier impact depths after the emplacement of a surface barrier. Four impact depth terms are used to convey the protective effect: drainage front, average drainage, the location with 50% impact, and drainage tail. The drainage front depth is the deepest point at which the barrier has a detectable impact at a specific time (also called the near zero-impact depth). The average-impact depth is the depth at which average drainage occurs. At the 50% impact depth, the water flux rate is reduced by half because of the surface barrier. Lastly, the drainage tail depth (also called the full-impact depth) is the deepest depth at which the water conditions above it are in equilibrium with the barrier. The algebraic expressions show that the average-impact depth is no more than 1/3 of near zero-impact depth, while the 50% impact depth is slightly larger than 1/2 of the near zero-impact depth. The full-impact depth, depending on the final recharge rate from the surface barrier, is usually much smaller than the other impact depths. These differences lead to a very large transition zone beneath a surface barrier. Numerical simulations were conducted to replicate the same conditions. The numerical results corroborated the analytical models by predicting very similar water content profiles and near zero-, average-, 50%, and full-impact depths. The algebraic expressions provided in this paper are useful for quickly identifying sites where the depth of the existing contaminants could be beyond the protection of a surface barrier.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Panel Session 79: Challenges and Lessons Learned in Radwaste Packaging, Transportation, Compliance, and Receipt

This panel focused on challenges and lessons learned in radioactive waste packaging, transportation, compliance, and receipt for fabricators of radioactive waste packaging. World-wide companies are urged to participate. Participants provided their perspectives and suggestions on how to increase efficiencies through eliminating inconsistencies between specifications and other requirements to provide products at a reasonable cost and enhance communication to enable better service. Panelists with presentations: PACTEC (Bill Smart); Waste Packaging Strategy and Challenges at Whiteshell (Miles Smith); Challenges and Lessons Learned in Radwaste Packaging, Transportation, Compliance and Receipt (Dan Payne)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Conceptual Design for a Short-Lived Low- and Intermediate-Level Waste Repository at the Al Tuwaitha Nuclear Center, Baghdad, Iraq - 17189

The European Commission under their Instrument for Nuclear Safety Cooperation (INSC) has initiated a project supporting Iraq's efforts for the safe disposal of short lived low- and intermediate-level radioactive waste (SL-LILW) arising from the decommissioning of Iraqi nuclear installations. The INSC project includes both the development of an initial conceptual level basic design and preliminary safety assessment (PSAR) report and the subsequent detailed engineering design and supporting intermediate safety assessment report (ISAR). The project is being executed by a Consortium led by NUKEM Technologies Engineering Services GmbH and joined by DBE Technology GmbH of Germany; JAVYS a.s. of Slovakia and MCM Environmental Services Ltd of the United Kingdom. The current paper focuses on the development of the conceptual level basic design. In previous decades Iraq had a significant nuclear program which encompassed ten nuclear sites spread out across the country. The largest site was located at the Al-Tuwaitha site near Baghdad where 18 nuclear facilities were located. The facilities suffered significant damage from bombing and subsequent looting during the first Gulf War in 1991. Currently, the Ministry of Science and Technology (MoST) is the radioactive waste management operator in Iraq and also the license holder for the ten Iraqi nuclear sites, all of them heavily damaged. Because of the ongoing security concerns in Iraq MoST has selected the Al Tuwaitha Nuclear Center as the preferred and best suited location for development of an SL-LILW repository. As an initial step in development of the conceptual design various repository design options for the site were examined and assessed. The assessment considered similar facilities that have been implemented internationally. At the conclusion of the assessment and in conjunction with the Iraqi Beneficiaries from MoST, a near-surface vault design was selected for the facility. The design of the facility relies on experience gained from existing low-level waste repositories at Centre de l'Aube in France and the El Cabril solid radioactive waste disposal facility in Spain as well as on more recent experience gained in the design of the Bulgaria National Disposal Facility (NDF) under construction near the Kozloduy Nuclear Power Plant. Similar to these reference designs, the concept for the Al Tuwaitha Disposal Facility (ATDF) will rely on a series of reinforced concrete vaults into which containerised waste will be placed. Each vault will consist of two chambers designed specifically to hold the waste. Central drainage points in each waste chamber will lead to an infiltration water collection system designed to detect any failures in the waste containment. Each disposal vault will be sealed after filling by installation of a reinforced concrete closure slab. Once all of the waste has been disposed of the site will be decommissioned and a multilayer protective cover will be installed over the sealed disposal vaults. The multilayer cover will act to further isolate the waste from the environment by limiting the access of water to the waste. Due in part to the presence of shallow groundwater at the site and to avoid potential damage to the facility by flooding from the nearby Tigris River the concept for the Al Tuwaitha Disposal Facility requires construction of the disposal vaults on a raised platform several meters above ground level. Construction of this platform will be integrated into the conceptual safety design and materials will be selected which further enhance the isolation function of the disposal system by hindering or delaying the release of radionuclides to the environment. The infiltration water collection system, which will be constructed in a specifically designed accessible gallery embedded and fully integrated into the platform, will provide access to inspection points located beneath each disposal chamber. The ATDF when completed will be a further demonstration of the adaptability of the disposal vault concept for use in isolating SL-LILW from the environment. (authors)

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DOE OSTI2017

Optimization of Low Level Waste Disposal Policy, Strategies and Techniques - 17201

The mission of low level radioactive waste disposal facilities is to accept and dispose of waste in compliance with all applicable laws, regulations, license requirements, and social expectations. These facilities need to provide a cost effective disposal service to the waste generators, so that the cost savings could be passed down to business, industry, and ultimately, to the general public. To fulfil this mission, waste disposal facilities need to establish and maintain review and optimization mechanisms to respond to changing circumstances that may impact the long-term safe, cost effective operation of their facilities. Maintaining existing disposal resources available as long as possible, managing increased demand for disposal space as a consequence of growing number of nuclear facilities to be decommissioned, and developing capacity for future needs and uncertainties are some of the key challenges that drive optimization of disposal facilities. While the concept of optimization of low level waste disposal facilities is not new, the framework for holistic optimization of waste disposal facilities, enabling them to provide cost effective, sustainable services to the nuclear industry, as well as maintaining optimized safety and radiation protection systems is a notable gap in the current radioactive waste management body of knowledge. A comprehensive optimization framework is required to support the implementation of safe, effective, and sustainable integrated waste management strategies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of Waste Inventory for Probabilistic Performance Assessment Modeling - 19361

Estimation of radioactive waste inventory is a key initial step in the development of an environmental fate and transport model for performance assessment (PA) of radioactive waste sites. PA model output is often sensitive to parameters that represent inventory as a source term, yet estimation is a difficult problem. Inventory estimation for a probabilistic performance assessment (PPA) is related to, but distinct from, that performed for waste excavation, characterization, and offsite disposal. The latter appropriately focuses on obtaining the most detailed information possible on specific hazards and locations; however, these questions cannot be fully answered without intrusive site characterization efforts. This type of characterization is often not feasible when assembling information regarding a source term in a PPA model. Development of waste inventory for a PPA is one piece of a larger hierarchical model. The physical process model represented by Goldsim computer code represents one level of the hierarchy, and the second level of the hierarchy involves estimation of inputs such as inventory, transport, and risk parameters. The estimation of inputs is not a straightforward problem. The parameters of a computer model often do not have an equivalence to the physical quantities they describe, and estimation of the physical quantities themselves can be difficult due to limitations of available information. Inventory information for the West Valley Site in upstate New York is based on data from historical shipping and disposal forms. In the 1990's, the West Valley Demonstration Project (WVDP) compiled information from existing databases and historical waste records into one complete modern database. A conceptual model was then applied and used with recorded information to characterize the waste mass and radionuclide composition for each waste burial. Information was ultimately aggregated over all waste records to estimate the mass of each radionuclide within each trench, hole, or caisson. Three decades later, we revisit the master database, make updates with newly-identified information, and incorporate the conceptual model previously used for inventory estimation in the hierarchical modeling framework. We characterize uncertainty using probability distributions for waste mass and radionuclide composition at the individual shipment or individual burial level. Uncertainty incorporated at the lowest level of the hierarchy is ultimately carried through to uncertainty in PPA model endpoints. Sensitivity analysis of early model results is then applied to identify influential inputs and guide subsequent efforts towards parameters that drive risk. (authors)

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