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DOE OSTI2015

SORTIE DB - Source term integrated environmental management disposal facility database - 15249

The Environmental Management Disposal Facility (EMDF) in Oak Ridge, Tennessee, is a proposed U.S. Department of Energy low-level Comprehensive Environmental Response Compensation and Liability Act (CERCLA) waste disposal facility with an operational start in the early 2020's. As a DOE low-level waste disposal facility, it is subject to DOE Order 435.1, requiring a composite analysis and possibly a performance assessment. The Source Term Integrated EMDF Data Base - SORTIE DB - provides source term estimates used in these assessments. An Oak Ridge Office of Environmental Management (OREM) Federal and Contractor team developed this state-of-the-art descriptive analytics and visualization schema. DOE OREM senior management has recognized complementary beneficial uses of SORTIE DB in decisions for accelerating clean-up progress, dealing with conflicting sequencing and budget priorities, and enhancing stakeholder and regulatory interfaces. SORTIE DB is designated as an OREM planning tool. SORTIE DB contains estimated forecast waste volumes and masses for over 800 Decontamination and Decommissioning (D and D) and Remedial Action (RA) sequenced baseline packages over the next 25+ years. Available characterization data, radionuclide identification and concentrations, and data from safety basis and emergency management documents are incorporated into SORTIE DB. (authors)

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DOE OSTI2015

Planning large-scale excavation and construction activities within an aerially constrained site for construction of a repository for low and short-lived intermediate level radioactive waste at the Bulgarian national disposal facility - 15372

The Bulgarian State Enterprise for Radioactive Waste Management (SERAW) is planning the construction of a near-surface repository for low and short-lived intermediate level radioactive waste, the National Disposal Facility (NDF), at the Radiana Site adjacent to the Kozloduy Nuclear Power Plant (KNPP). The Site, a rectangular area approximately 1.2 km east-west by 0.4 north-south, is located on a slopping area between the second and sixth loess terraces above the River Danube with elevations ranging from approximately +40 m to elevation +95 m. Construction of the NDF will be divided into three phases separated by approximately 20 years each. The first phase will see the construction of the primary support facilities, referred to as the Auxiliary Buildings and including the Waste Reception and Buffer Storage (WRBS) Building, and the first set of 22 reinforced concrete disposal cells and supporting infrastructure, including a network of subsurface observation galleries, the Infiltration Network Control (ICN) Galleries. Each disposal cell will be 20.15 m long by 17.05 m wide by 9.45 m in height. The disposal cells will be arranged in two parallel rows. Disposal activities will be conducted using overhead cranes installed inside of track mounted mobile roofs. Each row of disposal cells will be serviced by one mobile roof. The mobile roofs will be supported by massive concrete foundations. Construction at the site will present several logistical challenges that will need to be overcome to fully realize the NDF. First in order to begin construction of these facilities nearly 1 million 3 of soil will require excavation and handling. Consistent with requirements to minimize potential environmental impacts identified in the site Environmental Safety Assessment (EIA) excavated soil will, to the extent possible, remain onsite. Approximately three quarters of the soil can be accommodated within the site boundaries. Another challenge results from requirements imposed by the ground conditions. The sub-grade at the site, which consists primarily of loess, will require considerable improvement. To this end a 5 m thick loess-cement cushion will be constructed across much of the site. Specifically, the cushion will be required beneath the disposal cells and the WRBS Building. Construction of the loess-cement cushion in the disposal area will require detailed planning and coordination, particularly as the cushion beneath the disposal cells also contains embedded structures, including the ICN Galleries as well as massive concrete foundation beams required for the two mobile roofs. Construction of these structures will require planned interruptions in the construction of the loess-cement cushion. Specific cement types and strict quality requirements will be required for completion of the loess-cement cushion as well as for completion of the ICN Galleries. The construction planning will need to account for the large volumes of soil to be excavated as well as ensuring sufficient material supplies in terms of both quality as well as quantity. Large quantities of materials including both dry cement for mixing with the loess, and wet concrete for pouring of the mobile roof foundations are accounted for by the plan. However, even after completion of these activities additional challenges are addressed by the plan related to the contemporaneous construction management of the auxiliary buildings and the disposal cells. (authors)

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DOE OSTI2015

Salt Waste Processing Facility project lessons learned: Acquisition through construction - 15518

The Salt Waste Processing Facility (SWPF), located at the Savannah River Site (SRS) in Aiken, South Carolina, is a multi-billion dollar Liquid Radioactive Waste (LRW) processing facility that represents the keystone for the U.S. Department of Energy (DOE)-EM clean-up mission at SRS. The SWPF facilitates DOE-EM clean-up at SRS by decontaminating the salt waste, which is the dominant volumetric component in the LRW tanks, thereby allowing the cleaned salt waste to be removed from the tanks and stabilized into an approved grout-based final waste form. The small volume of removed radionuclides is sent to the Defense Waste Processing Facility (DWPF) for vitrification. Emptied LRW storage tanks are then grouted into an approved configuration for final closure. The SWPF project is currently 100% complete with design, 80% complete with construction, and 10% complete with commissioning. To date there have been numerous lessons learned with regard to project acquisition, design, procurement, construction, and quality assurance (QA) approaches (both good practices and mistakes) that are considered to be of significant value in improving the state of major project execution within the DOE-EM clean-up program as well as to the wider DOE enterprise. An overview of the current SWPF project status is provided and key lessons learned in the aforementioned areas are presented relative to improving future DOE major project execution performance. Specific areas to be addressed include strategies to optimize integration of design and nuclear safety, constructability tools for design and construction synchronization, commercial grade dedication (CGD) approaches to mitigate atrophy in the nuclear supply chain, and QA perspectives to facilitate effective and efficient project execution. (authors)

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DOE OSTI2016

Development and Expansion of Research Facilities to Address the Challenges of the Nuclear Industry - 16055

With the wide variety of challenges that surround nuclear decommissioning, bespoke facilities are required to undertake experiments which can inform decision making. Here we describe research carried out at The University of Manchester's Dalton Cumbrian Facility, a custom built research environment which incorporates a 5 MV ion accelerator as well as a self-shielded {sup 60}Co irradiator. The ion accelerator allows the investigation in to the radiolytic consequences of various charged particles, including protons, alpha particles and a variety of heavier (metal and nonmetal) ions, while the {sup 60}Co irradiator allows the effects of gamma radiation to be studied. We provide specific research examples that demonstrate how this equipment can improve our mechanistic understanding of various aspects of the deleterious effects of radiation in the nuclear industry. These examples include applications in waste encapsulation as well as geological disposal and novel surveying techniques. The outlook for future research and expansion of the facility is also presented. (authors)

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DOE OSTI2016

Facility Improvements and Modifications to Extend the Operational Life of the Modular Caustic-Side Solvent Extraction Process - 16082

Savannah River Remediation, LLC (SRR) is working to remove, stabilize, and dispose of approximately 36.7 million gallons (138.9 million liters) of liquid radioactive waste currently stored in 43 underground waste tanks, and ultimately close waste tanks. In 2008, as an interim salt treatment system to remove cesium from high level waste salt solutions prior to the Salt Waste Processing Facility start up, the Modular Caustic-Side Solvent Extraction Unit (MCU) initiated the removal of cesium from liquid radioactive salt waste prior to disposal. The MCU was designed and built for an expected three-year operational life and five-year design life. The first MCU cold runs began in June 2007 and hot startup was completed in April 2008. MCU remains in hot operation to this day, aiding in multiple, high-level tank closures. In an effort to continue liquid waste disposition at the Savannah River Site (SRS), equipment upgrades and operational improvements have been made to extend the life of the MCU beyond the original three years. Several pieces of fixed equipment have been converted into removable equipment which reduces the downtime associated with equipment change-outs and thereby reduces the dose to workers and processing impacts. Multiple pumps and filters have been upgraded to more robust designs. Process monitoring and controls have been upgraded to improve process performance. Redundant centrifugal contactor vibration equipment has been installed on each contactor in order to better assess the reliability, processing performance, and lifetime of the contactors. A new Solvent Hold Tank (SHT) Sampler has been installed to minimize the organic cross-contamination within the sampler, dose to workers, and sample analysis time. To increase processing availability and aid in visually obtained data, camera port has been added to the Salt Solution Feed Tank (SSFT). In order to decrease the overall background dose to the workers, the Process Vessel Ventilation (PVV) duct work now has the capability to be flushed to remove contamination. The caustic wash addition system has been modified to be a once-through system to improve equipment reliability, decrease downtime, and decrease dose to workers. In addition, the solvent that MCU utilizes to extract Cesium out of salt waste has been upgraded to Next Generation Solvent (NGS) which is based on MaxCalix extractant instead of a BobCalix extractant. Further, these upgrades and improvements have allowed MCU to be able to process at higher flow rates (up to 8.5 GPM) and with increased throughput (as high as 1.3 million gallons in FY13 and 5.3 million gallons since hot operations initiated) that have allowed SRR to continue the disposition of nuclear salt waste. The continued operation of MCU ensures more curies are stabilized into a glass matrix at SRS's Defense Waste Processing Facility (DWPF) and less curies are retained within South Carolina at SRS's Saltstone Disposal Facility (SDF). Unique challenges associated with modifying and improving a demonstration plant that is in hot operations and was designed to be temporary with a limited plant footprint have been successfully overcome. Further, creative strategies for removing old equipment have been utilized given the limited decontamination agents available for use that will not interfere with the organic solvent. This paper presents and discusses the deployment of these upgrades and operational improvements to MCU in order to allow continued, increased salt waste processing beyond the initial design life. (authors)

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DOE OSTI2016

Life Extension of Nuclear Plant Facilities through Monitoring and Detailed Analysis - 16175

The UK has accumulated a substantial legacy of radioactive waste from various civil nuclear and defence programmes. Presently, however most of the UK radioactive waste is generated from nine power stations which generate approximately one sixth of the of the country's electricity. Irradiated fuel is removed from these reactors and safe management of the radioactive solid and liquid effluents is required. The safe management of these effluents is essential. Sellafield, in West Cumbria UK, contains two reprocessing plants, one designed to process nuclear fuel from Magnox reactors (CO{sub 2} cooled with Uranium bar fuel) and the other to process nuclear fuel from Advanced Gas cooled Reactors and Pressurized Water Reactors. The reprocessing stage enables extraction of fissile material for recycling. This leaves a volume of dilute but high-level waste. The high-level radioactive liquid waste is concentrated to reduce its volume and then incorporated into borosilicate glass, using a process called vitrification. The glass is then poured into stainless steel containers and placed in long term storage. This required concentration of HLW liquor is achieved using a suite of aging evaporators. All evaporators are of similar design constructed from corrosion resistant stainless steel having dimensions height ∼11 m, internal diameter 3.05 m and capable of holding 13.7 m{sup 3} of liquor. The basic principle of operation is to heat the dilute liquor under partial vacuum, at constant volume, until the target concentration is achieved. Heating of the liquor is achieved using steam heated coils within the evaporator, and steam in the jacket which surrounds the lower half of the evaporator vessel. The evaporator is then brought to atmospheric pressure and the concentrated liquor is ejected whilst hot. The evaporator is then cooled with water cooled coils and a water filled jacket. New dilute liquor is then added for a new batch. The process is then repeated. As the evaporators reach the end of their original design life accurate prediction of the current condition of nuclear plant facilities is essential to safely justify continued plant operation by ensuring structural integrity. This poses a significant challenge as many aging UK nuclear facilities were not inherently built with inspection in mind. The potential life limiting components of the HA evaporators at Sellafield are the surfaces that contact the product liquor at elevated temperatures i.e. the steam heated surfaces of the coils within the evaporator vessel and the vessel base and wall surrounded by the steam jacket. Each heating component is critical to the operational life of the evaporator - the removal from service of a coil heating component significantly reduces the operational flexibility and remnant life of this plant. Over the last decade a variety of cutting-edge expertise and methodologies have been utilised to accurately measure and then predict the current condition of the plant and also to predict its remnant life. This paper provides an overview of the development of several techniques deployed, namely non-destructive testing/inspection, in-house software development and finite element modelling. These methods have been validated by lab-scale or full-scale experiments along with historical plant data. These techniques have been evolved and refined for over a decade, providing vital insights into the physical behaviour of this complex reprocessing system. This enables not only accurate assessment of the current condition of the plant equipment but also allows scientifically underpinned predictions to be made of how it will behave into the future. This is critical for Sellafield as this knowledge feeds into their safety case to the Office of Nuclear Regulation to justify continued operation. It is concluded that this integrated approach provides a robust analysis tool which has driven change and optimisation of plant procedures ensuring continued safe operation whilst extending the life of the facility. (authors)

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DOE OSTI2016

Large Cask, Small Building, AREVA's Safe MLLW Disposal Solution Despite Facility Limitations - 16295

In order to begin decommissioning of a 1950's research reactor, Class C Mixed Low Level Waste (MLLW) had to be removed from the facility and disposed of. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to the building layout, crane capacity and pool size, the standard TN RAM cask loading process could not be utilized. Unique solutions and approaches had to be developed to complete the project. AREVA designed a smaller liner insert that would house the four MLLW components in a single waste package to solve the problems related to crane capacity and physical pool dimensions. AREVA also designed and fabricated a shield bell to provide radiation shielding while moving the liner insert from the reactor pool to the cask that was staged adjacent to the reactor pool. The AREVA team worked with the disposal facility to cover requirements for disposal of Class B MLLW. This included designing spacers filled with inert material to fulfill the requirement for maximum 10% void space in the waste package. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for disposing MLLW requiring macro-encapsulation in a single waste package. This solution also provides a model for how to adapt large shipping casks and liners to smaller aging commercial, DOE or research facilities. (authors)

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DOE OSTI2016

Low Activity Waste Glass Waste Form Data Package for Hanford's Integrated Disposal Facility Performance Assessment - 16480

A Direct Feed Low Activity Waste (DFLAW) flowsheet has been initiated to begin immobilization of Hanford tank wastes. In the DFLAW configuration, a pretreatment system will segregate the high level and the low activity wastes (LAW). The LAW will be immobilized in the Waste Treatment Plant (WTP) and disposed of in Hanford's Integrated Disposal Facility (IDF). To obtain authorization to dispose of wastes in the IDF, a Performance Assessment (PA) must be completed in accordance with DOE Order 435.1. The PA shall include calculations of potential doses to representative future members of the public and potential releases from the facility to provide a reasonable expectation that performance objectives will not be exceeded as a result of operation and closure of the facility. To support the development of the PA, data from ILAW glass testing are needed to model the long term release of contaminants over the compliance period and verify that impacts to groundwater quality will be within performance objective limits. Washington River Protection Solutions is conducting an ILAW glass testing program to produce the required data. This program is a combined effort involving National Laboratories with known expertise in glass waste form research. The work involves accelerated testing to develop parameters that can be used in computer simulations of glass corrosion and contaminant release in the IDF environment. The results of this work will be compiled in an ILAW glass waste form performance data package for the IDF PA. (authors)

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DOE OSTI2018

Regulating Operational Safety at a Future Geological Disposal Facility in the UK - 18038

The United Kingdom (UK) has generated commercial nuclear energy for over 60 years; generating a significant and varied inventory of higher activity radioactive waste (HAW). In accordance with international practice, UK Government policy for long-term management of HAW is disposal in a geological disposal facility (GDF). The nuclear materials comprising HAW include high level waste, intermediate level waste and some low level waste, not suitable for disposal to the Low Level Waste Repository. In addition, other nuclear materials may be reclassified as waste in the future and will require to be managed as HAW for disposal to a GDF; these include spent fuel, uranium and plutonium. The operator of a GDF will be required to obtain a nuclear site licence from the Office for Nuclear Regulation (ONR), as the nuclear safety and security regulator, in order to construct and operate the facility. In the period prior to submission of a licence application, ONR is working to develop its regulatory expectations and provide advice on its regulatory role with regards to a future GDF. ONR has developed regulatory guidance on operational safety, benchmarked against the relevant IAEA Safety Standards and the Western European Nuclear Regulators Association (WENRA) Safety Reference Levels for Disposal. This guidance draws upon ONR's current expectations for nuclear sites; including operational reactors, fuel manufacture and reprocessing facilities, and reactors undergoing decommissioning; and incorporates relevant good practice drawn from international programmes for geological disposal and related industries (eg mining). In addition to the technical assessment guidance, ONR has developed guidance on the licensing process specific to a GDF; providing clarity for example on when a licence will be required. A review of the standard licence conditions applied to all nuclear sites concluded no significant amendment would be required for application to a GDF. Through its voluntary agreement with Radioactive Waste Management Limited (RWM), the developer and future operator of a GDF in the UK, ONR exerts regulatory influence on the design and organisational development of the future licensee. As part of this scrutiny, ONR has assessed RWM's generic Disposal System Safety Case. Subjecting the generic design and safety case to regulatory review provides opportunity for early engagement to ensure the appropriate standards will be achieved. ONR will only grant the necessary regulatory permissions if RWM adequately demonstrates the GDF will be constructed, operated and closed safely. Future challenges faced by ONR with respect to a GDF include building and maintaining regulatory competence and confidence, and engaging effectively with the public as the UK begins its search for a suitable site to provide appropriate assurance that a GDF will be operated safely and securely. (authors)

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DOE OSTI2018

A Novel Mobile Examination and Remediation Facility for On-Site Remediation of Dry Storage Systems - 18389

Most of the spent nuclear fuel (SNF) in dry storage in the United States is safely and securely stored in dual-purpose (i.e., storage and transportation) canisters [1]. However, some of the SNF is stored in storage-only (i.e., non-transportable) packages, or packages that may become nonconforming for continued storage or transportation for a variety of reasons. Currently, the capability to remediate dry storage systems (DSSs) for SNF that are not transportable or that need to be repaired to maintain confinement boundary integrity during storage service does not exist without on-site access to a spent fuel pool. Hence, there is a potential future need for a mobile facility with remediation capabilities that can be used at multiple storage installations. These capabilities would permit needed repairs to an existing DSS to support continued storage or for the dry transfer of fuel to a new conveyance suitable for transport, if necessary. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. A novel design concept for a Mobile Examination and Remediation Facility (MERF) is proposed with functions and capabilities to remediate a DSS. The proposed MERF would be a multi-purpose facility, remotely operated and computer-controlled with semi-automated equipment to allow trained operators to safely perform remediation activities while keeping worker radiation dose as low as reasonably achievable. It would include a shielded rotating platform with support towers and would use commercial off- the-shelf examination and repair equipment. A slip ring arrangement underneath the rotating platform would allow for the power supply without impeding the continuous rotation and would simplify cable management. The support towers would be mounted on the rotating platform, and each tower would have vertical and horizontal positioning capability for axial and radial tool location. The rotating platform would be attached to the main MERF structure. Capabilities would include visual, surface, and volumetric examinations to fully characterize physical conditions on canister surface before and after a repair; repair techniques, e.g., laser beam welding, friction stir welding, or cold spray, to perform needed remediation activities; and associated equipment for surface preparation such as cutting and grinding equipment. The proposed design would be simple, modular, robust, economically viable, and easily transportable and would have a small footprint. The modular design would present several advantages including compatibility with a variety of different DSSs, on-site assembly, and transportability; it also would be built in interchangeable sections to facilitate decontamination. Further, shielding would be supplemented through interface with the existing transfer cask and storage overpack. Robustness and canister stability would be achieved using pneumatic holders and steel support structures designed with vibration absorbers to reduce impacts and/or a 'pendulum' effect during a seismic event. Thus, the proposed MERF would be designed to: (i) minimize handling of an affected DSS to the extent practicable, (ii) provide radiation shielding and minimize dose to operators and the public during remediation activities, and (iii) enable examination, repair, and repackaging. This paper describes the proposed MERF design concept and capabilities, main components, equipment, different modes of operation, and operating procedures. Structural and safety assessments would need to be performed before final design, licensing, and deployment. The MERF is evaluated with respect to regulatory impact (first of a kind, prior experience), operational impact (safety, security, staffing, ongoing operations, constraints), worker dose (low, medium, high), cost (low, medium, high), schedule (duration, time sequence), and technology readiness level. (authors)

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DOE OSTI2018

Cold Commissioning Simulant Selection for the Hanford Tank Waste Treatment Plant LAW Facility - 18465

Commissioning the Low-Activity Waste (LAW) Facility of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) will require large volumes (∼2,000,000 liters) of a non-radioactive, chemically analogous waste simulant. Cold commissioning will provide for operator training in an integrated facility, demonstration of mechanical process equipment capacity, environmental performance demonstration of emissions control systems, and process control systems shakedown and tuning. Selection of the cold commissioning simulant is following a structured four-phase process to arrive at the preferred composition and recipe. Phases I, II and III are complete. Phase I laid the framework for decision making with preparation of a Project Management Plan, an update to the Simulant Development, Approval, Validation, and Documentation procedure, and a kickoff meeting to gather stakeholder input into the simulant definition and selection criteria. Phase II consisted of a comprehensive literature search and compilation of LAW simulant recipes, followed by initial screening against the selection criteria identified in Phase I. This process identified 72 candidate simulant recipes and down-selected 9 for further study. Phase III studies involved a) statistical comparisons of the similarity of candidate simulant compositions to the tank waste feed vector, b) process modelling of the commissioning sequence with each simulant to estimate glass former and simulant material consumption, offgas and effluent volumes and compositions, and product glass volumes and compositions, and c) evaluation of simulants against the immobilized LAW glass composition algorithm to determine allo able waste loading to ensure compliance with product specifications and process constraints. The results of these studies identified two of the candidate recipes to be carried forward to full-scale testing. Phase IV consists of testing these two recipes at vendor facilities to determine scalability, fabricability, and stability of the simulants when produced at industrial scale. Scalability is the ability to take a recipe from bench scale (1 to 40 liters) batch size to full scale (>35,000 liters) batch size. Fabricability is the ability of a recipe to be prepared in large batches without the need for complex or expensive mechanical equipment, heating, or cooling systems. Stability is the ability of a simulant to withstand long term storage and variations in environmental conditions without unacceptable changes in key characteristics (the set of chemical, physical, and rheological parameters that the simulant is intended to exhibit during cold commissioning tests). Simulants will also be tested for compatibility with other startup material (i.e., frit), tuning simulants, and environmental performance demonstration test spiking solutions to ensure against unexpected situations. Using a structured process is essential when making high impact decisions such as waste simulant selection. Proper planning and a clear explanation of how the decision process will unfold is key to obtaining stakeholder buy-in as more information is gathered. The documentation resulting from this structured selection process will be compiled and summarized in a Cold Commissioning Simulant Definition deliverable due to the Department of Energy 24 months before the start of Cold Commissioning. (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 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)

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DOE OSTI2019

Operational Safety Analysis for Disposal Period of the Future Belgian Geological Disposal Facility - 19249

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth between 200 m and 600 m. ONDRAF/NIRAS has developed a reference disposal concept for research, development and demonstration (RD and D) purposes. At the current stage of this program, the primary objective of ONDRAF/NIRAS is to establish the safety and feasibility of constructing, operating and closing a Geological Disposal Facility (GDF) at the conceptual design level, without any presumption about the exact location of the facility. This will be evaluated in a first version of the safety and feasibility case (SFC-1 V1) for the geological disposal of category B and C radioactive wastes. Analysis of the operational safety is an important part of the SFC-1 V1. ONDRAF/NIRAS, supported by BGE Technology GmbH, has prepared an operational safety analysis of the current GDF reference design and the processes occurring during the operation and closure period. A SWIFT methodology (Structured What If Technique) is used to identify risks and the measures required to mitigate their consequences. In parallel to this SWIFT analysis, specific attention is also given to fire risk, especially in the blind-ending disposal galleries (DG). Based on understanding of fire risk analysis, recommendations suggest using exclusively electrically driven equipment to reduce fire loads and minimize the fire hazard. In addition, strict working procedures could prevent the presence of workers between the DG dead end and equipment presenting significant fire load. The use of mobile refuge chambers should also be part of the safety concept. (authors)

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DOE OSTI2019

Performance Assessment for Closure of the Calcined Solids Storage Facility at the INL Site - 19289

The Calcined Solids Storage Facility is composed of six discrete stainless steel 'bin sets' at the Idaho Nuclear Technology and Engineering Center at the INL Site. Approximately 4,440 m{sup 3} of the dry, granular calcine waste is stored in the facility. Closure of the facility will entail removal of the calcine waste followed by grouting. A 5-cm thick residual amount of calcine was assumed to remain at the bottom of the bins. Important radionuclides in the residual layer for the groundwater pathway were identified as the fission and activation products Tc-99, Se-79, I-129, Cs-135, H-3, and actinides Pu-239, and Pu-240. The stainless-steel bins prevent near-term release of grouted residual calcine waste to the environment. Future releases were conservatively based on the corrosion rate of the weld seams on the stainless-steel bins, which site-specific studies measured mean tunnel corrosion rates of 0.0132 mm/yr with a coefficient of variation of 0.333. Based on tunnel corrosion rates, mean bin hydrologic containment times were estimated to be between 238 years and 1,191 years, depending on the thickness of the bin. Any breach in the weld seam was assumed to result in water flow and radionuclide leaching through the bin. The structural integrity of the bin was assumed to be controlled by the general stainless-steel corrosion rate, which was much slower (3.5 x 10{sup -5} mm/yr), resulting in structural integrity times between 24,000 years to over 100,000 years. A mixing-cell model was used to compute water flow and transport through the hydrologically failed bins and the vadose zone. Credit was taken for the geochemical properties of the grout used to fill the bins at closure, which significantly slows the releases from the bins for key radionuclides like Tc-99. A semi-analytical model for three-dimensional dissolved-phase transport in an aquifer of constant unidirectional flow, infinite length and width, and finite thickness was used to compute concentrations in the aquifer across the model domain for a time varying source. Cumulative impacts from all bin sets were calculated by summation across the bin set sources that were superimposed on a uniform Cartesian grid. Effective doses were calculated for direct groundwater ingestion from wells located 100-m downgradient from each of the bin sets and ingestion of meat, milk, and vegetables irrigated with water derived from the well. Intruder doses were based on a well drilled through the grouted waste and cuttings spread around a home site after a 500-year institutional control period. Peak doses during the 1,000-year compliance period were less than 5% of the 250 μSv/yr (25 mrem/yr) performance objective and intruder doses were generally less 5% of the of the 5000 μSv/yr (500 mrem) and 1000 μSv/yr (100 mrem) performance objective for acute and chronic intruder exposure, respectively. Groundwater concentrations were all less than the maximum contaminant limits. In all cases, doses were highest for the oldest bin set because, compared to the other bin sets, it had the highest radionuclide concentrations in the calcine waste. The performance assessment results indicate that performance objectives can be met leaving considerably more than 5-cm of residual calcine waste in the bin sets. (authors)

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DOE OSTI2019

Advanced Modeling of Contaminant Fate and Transport in Soil and Groundwater from F-Canyon and the Central Laboratory Facility

F-Canyon (221-F) was constructed in the early 1950's and, together with FB Line, began operations in 1954 as the main processing facilities in FArea of SRS. The primary operations conducted in F-Canyon and FB Line included the separation and recovery of Pu-239 and U-238 from irradiated materials and the stabilization of plutonium residues. The F/H Laboratory Complex consists of buildings 772-F, 772-1F, and 772-4F which began operations in 1954, 1987, and 1993, respectively. The primary mission of the F/H Laboratory Complex for the last 45 years has been to support the chemical separations processing activities at Building 221-F and H-Canyon. The objective of this study is to determine the radionuclide and hazardous material holdup in F-Canyon and its support facilities, and to evaluate In-Situ Disposal (ISD) alternatives for deactivation and decommissioning (D and D) of these facilities to ensure groundwater protection and environmental compliance with the South Carolina Department of Health and Environmental Services (SCDHES) and the United States Environmental Protection Agency (USEPA). GoldSim is a dynamic probabilistic simulation software that allows modeling of chemical mass transport through engineered and environmental systems For this study, it will model the potential future release of radiological contamination through the vadose zone, Upper Three Runs (UTR) Aquifer and the Upper Three Runs Creek. Using a GoldSim model developed for building 235-F D and D, contaminant inventories as well as building geometry were updated to create the basis of the new F-Canyon model. The inventories from F-Canyon, FB Line, and the 800 Series Tanks (Tanks 804, 808, and 809) as well as F/H labs (including buildings 772-F, 772-1F, and 772- 4F) were used to make a total of four sources. The concentrations of total uranium and beta-gamma are within compliance for the study, while radium and alpha concentrations exceed their MCL's around 30,000 years and total effective water dose exceeds its MCL around 1,200 years. This suggests that some level of action will be needed to meet SCDHES and USEPA requirements. This evaluation was conducted only for the no-action alternative. Consideration and modeling of other alternatives will be required in the future to determine the best ISD alternative for D and D of F-Canyon and F/H labs with regard to groundwater protection, worker safety, and financial cost.

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DOE OSTI2020

Panel Session 46: Final Closure of Unlined LLW Disposal Facilities Worldwide (R3.5)

This panel discussed the on-going and emerging issues with the design and construction of final cover system of unlined disposal facilities (e.g., material disposal areas (MDAs), subsurface disposal areas (SDAs), burial grounds, and similar historical waste trenches, pits and vaults). The discussion focused on key technical and regulatory issues as well as lessons learned from years of engineering and construction in US and non-US locations. Panelists with presentations: Closure of Unlined LLW Disposal Sites (Craig Benson); Closure of LLW Facilities (Paul Pointon); Final Closure of Unlined LLW Disposal Facilities - Lessons Learned (Kevin Foye)

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DOE OSTI2020

Operations and Recent Upgrades to a Tritium Recycling Facility - 20375

Nuclear Sources and Services Inc. (NSSI) is a waste handling treatment company that has been operating since 1971. In 2000, NSSI began building a 1,000 m{sup 2} (∼10,000 sqft) tritium recycling facility on the Houston, Texas premises to address a growing need in the waste disposal market. Over the years, the capability of the facility grew to address many different waste streams with the goal of recycling > 99.9% of all tritium activity back into the fusion community. This is achieved by maintaining an effluent stack release well below the 150 GBq/year (C Ci/year) operating limit and maintaining all liquid and solid waste disposal below the low-level radioactive release limit of 0.C MBq/kg (10 μCi/kg). The unique processing systems in the facility attracts many challenging contracts. Within the past two years, NSSI has engaged with customers to treat; organic mixed wastes that simultaneously contain tritium and carbon-14, safely extract tritium from large quantities of exit sign lights, and remove tritium from uncharacterised legacy depleted uranium beds from a superfund site. This paper will outline the initial testing, the system upgrades to treat these varied streams, and the progress in processing the feedstocks provided by these projects. (authors)

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