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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)

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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)

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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 OSTI2015

A Methodology for defining intermediate level waste package characteristics that ensure criticality safety in a geological disposal facility - 14599

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. The basis for ensuring criticality safety of intermediate level waste (ILW) packages is to specify controls on the fissile material content of individual packages through the application of deterministic criticality safety assessments (CSAs). The approach is defined and controlled through RWM's Disposability Assessment process, which aims to ensure that: criticality is prevented during waste package transport to a GDF; the risk of criticality is tolerable and as low as reasonably practicable (ALARP) during GDF operations; and the likelihood and consequences of criticality arising from the evolution of waste packages after GDF closure are low. The most restrictive phase determines the package fissile material limit and, until now, in many cases the GDF post-closure phase has provided the most restrictive limit. Currently, the GDF post-closure limits are derived deterministically using highly stylised and conservative assumptions about fissile material accumulations occurring on very long timescales into the future. RWM, waste producers and regulators do not currently consider this approach to be yielding an appropriate balance of risk. Therefore, RWM is modifying the process for setting fissile material limits on ILW packages by making use of the outcomes of its recent research projects to establish the likelihood of post-closure criticality in a GDF and the consequences of hypothetical post-closure criticality events (presented in companion papers). This paper presents the revised methodology, which is based on the requirement that criticality in a GDF can be demonstrated to be of low and tolerable consequences with regard the performance of a GDF. A probabilistic 'low-likelihood package envelope' is proposed that establishes the packaging and disposal facility conditions under which post-closure criticality is considered to be of low likelihood and low consequence. The bounding conditions for establishing the extent of the package envelope have been evaluated through application of the methodologies and models developed in RWM's Likelihood of Criticality research project. It is expected that the vast majority of ILW packages would meet the package envelope requirements. The methodology also allows the use of other options, such as deterministic CSAs, for waste packages that fall outside the low-likelihood package envelope. (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)

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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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DOE OSTI2020

The Port Hope Areas Initiative's Port Granby Project: The Decontamination of Site Equipment from a Low Level Waste Facility - 20330

The Port Hope Area Initiative (PHAI) is the Government of Canada's response to the community recommended solutions for the remediation and safe, long-term management of historic low-level radioactive waste (LLRW) that was the result of uranium and radium processing operations in the Municipality of Port Hope, Ontario from the 1930's to the 1980's. The Eldorado refinery, on the shores of Lake Ontario, began refining radium-226 from pitchblende ore, later transitioning to the refining of uranium ore. Process residues were deposited at the Welcome Waste Management Facility in Port Hope until the mid-1950's switching to the Port Granby Waste Management Facility (PGWMF), which continued to accept waste until the late 1980's. The Port Granby Project involves the construction of a new, secure long-term waste management facility (LTWMF) and supporting infrastructure, and the excavation and relocation of approximately 1,200,000 m{sup 3} of historic low-level radioactive waste and marginally contaminated soil from the PGWMF to the newly constructed LTWMF. Since the project's inception in 2016, the bulk of the material has since been moved to the final long term storage cell. This has taken over 500,000 person-hours and a large amount of one site heavy equipment to accomplish including excavators, bulldozers, rock trucks, baker tanks and vacuum trucks. While many of the pieces of heavy equipment and water management equipment has further work to do on site, several of the sites larger items have already undergone the decontamination process prior to release from the site. This process has posed a serious challenge as the waste has a caustic nature to it and over the last few years it has had time to work its way into the metallic structures and framework of the equipment. Coupled with the tight work spaces and in some cases confined work areas there have been many issues that needed to be addressed to successfully reach a state where items can be free released to the unrestricted limits of the sites license. This ongoing process has led to a more refined decontamination program at the Port Granby Facility and has had a rather high success rate considering the length of time the equipment has been exposed to the waste materials on site. This paper discusses the varied challenges of decontamination faced at a low-level radioactive waste facility with a focus on the following aspects: - Challenges faced regarding different equipment types and the solutions implemented by Wood to address these challenges; - An examination of various decontamination methods employed throughout the project; and - Lessons learned over three years of remediation work and how they impact the decontamination process. (authors)

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DOE OSTI2017

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 17275

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DOE OSTI2018

The Status of LILW Complex Disposal Facility Construction in Korea-18511

This paper provides the status and experiences of Low and Intermediate-level radioactive Waste (LILW) disposal facility in Korea. The LILW disposal facility is being constructed in Gyeongju-si, Korea. It can dispose of 800,000 drums (hereafter, 'drums' means '200 liter drum equivalents' unless specified otherwise) of radioactive waste in a site of 2,060,000 m{sup 2}. The first stage of the LILW facility construction was completed in June 2014, and approved for use by Nuclear Safety and Security Commission (NSSC) in December 2014. The first stage of the LILW facility, which is a type of underground silo with disposal capacity of 35,200 m{sup 3} (100,000 drums), consists of a construction tunnel, an operation tunnel, an access shaft, and six silos, etc. Silo for final disposal is located 80∼130 m below sea level, 25 m in diameter and 50 m in height. The capacity of each silo is approximately 16,700 drums. The second stage of the LILW facility is under construction and will be completed by December 2020. The second stage of the LILW facility, which is a type of near-surface disposal with disposal capacity of 32,875 m{sup 3} (125,000 drums), consists of an underground gallery, three movable cranes and twenty vaults, etc. Vault for final disposal is located about 100 m above sea level, 20 m in width, 20 m length and 10.9 m in height. The capacity of each vault is approximately 6,250 drums. A second receipt and storage facility is to be completed by June 2021. This facility is under designing to prepare for further increasing radioactive waste receipt quantities. It has storage capacity of 10,000 drums (existing receipt and storage facility have storage capacity of 7,000 drums)and will be divided into several sections for pre-disposal work, storage, a passage to link the existing facility to the second receipt and storage facility and truck bay, etc. These components of the LILW disposal facilities are together called LILW complex disposal facility. It is the world's first LILW disposal facility with two different types of disposal facilities in one site. For efficient operation, these disposal facilities share a receipt and storage building, a waste water treatment system, a measurement and monitoring system, etc. Once the construction of the first two stages is completed, LILW disposal facilities will be incrementally expanded, and as the whole LILW disposal facilities are completed, all LILW in Korea will be disposed of these disposal facilities efficiently according to its radioactivity level and managed safely by KORAD. (authors)

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DOE OSTI2020

Declaration of the Chemical Composition of Waste-products for the German Repository Konrad Using Substance Vectors - 20500

The waste acceptance criteria of the German repository Konrad require beside a radiological declaration a compositional declaration to be able to avoid negative impact on the groundwater by a presumed leakage into the groundwater after complete corrosion of the containers in the future. The declaration of the composition has to specify compounds and substances not element analyses, because the chemical behavior and toxicity especially of organic compounds varies strongly The relevant compounds have to be described and quantified using defined substance codes provided by the operator. For simple waste types like pure concrete or steel components, the definition of the composition is rather simple. The composition of more complex waste streams is generally declared using producer specific material vectors, which can be derived for defined waste types (e.g. grouted fluids, ash from combustion, super-compacted mixed waste etc.). Additional substance vectors exist for containers, drums and grout, which together with the waste composition yield the total composition of the waste packages. The substance vectors for homogeneous waste streams like cementitious demolition waste, solidified fluids and ash are mostly based on analyses provided by the waste owner. In some cases, analyses from non-radioactive analogues or data sheets can also be used to demonstrate the composition. For each qualified treatment of radioactive waste streams the material composition is described and balanced on the basis of specific material vectors, which are composed of material codes from the material and cask list provided by the Federal company for radioactive waste disposal. For mixed waste streams like super-compacted or grouted mixed waste or materials simply packed in containers an analyses cannot be used. In these cases, the compositional information from the bookkeeping can be used for the entire waste stream to calculate a mean composition including a compositional range (min / max composition). Only waste packages with an approved material description can be delivered to Konrad. Therefore, the material declaration of all waste streams and the resulting waste products is crucial for all waste owners in Germany to be able to have enough waste packages ready, when the repository is going to open in 2027. (authors)

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DOE OSTI2016

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 16281

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to clean up 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. The WTP includes several facilities that will contribute to waste processing: (1) the Pretreatment (PT) Facility, which is designed to receive waste retrieved from the Hanford tank farms and separate it into low-activity and high-level waste streams; (2) the Low-Activity Waste (LAW) Facility, which will vitrify the low-activity waste stream; (3) the High-Level Waste (HLW) Facility, which will vitrify the high-level waste stream; (4) the Analytical Laboratory (Lab), which will support both vitrification facilities' operations; and (5) the Balance of Facilities (BOF), which will provide industrial support capabilities and infrastructure, like electricity and steam, needed for WTP operations. In 2012, DOE suspended production work on the WTP's Pretreatment (PT) Facility and to a lesser degree the High-Level Waste (HLW) Facility because of unresolved technical issues associated with those facilities. DOE has established an approach that would facilitate immobilization of tank waste to begin as early as practicable without waiting for completion of work to resolve the technical issues associated with the PT and HLW Facilities. This approach enables LAW treatment to begin before the PT and HLW facilities are completed by transferring low activity waste directly from the Hanford tank farms to the LAW Facility - referred to as Direct Feed LAW (DFLAW). ORP has placed the highest priority on completing the Low Activity Waste Facility, Balance of Facilities, and Analytical Laboratory (collectively known as LBL) in support of the planned phased start to the Hanford Tank waste treatment mission. DOE has requested, and the WTP contractor has delivered contract modifications proposals for: (1) design of a permanent capability to accommodate DFLAW in WTP; (2) procurement and construction of modifications to enable DFLAW operations in WTP, and (3) completion of LBL work scope in the current contract through hot commissioning in a manner to accommodate a phased start toward completing the WTP mission. DOE is reviewing and approving contract modifications in a sequence and schedule that allows for an orderly completion of the LBL Facilities and to begin treating low-activity tank waste as soon as practicable. In August 2014, ORP authorized the WTP contractor to resume all engineering work necessary to finalize the design of the HLW Facility. Authorization for HLW equipment procurements and construction is forthcoming, based on conditions that must be satisfied for a full release of these activities, including completion of an update to the Preliminary Documented Safety Analysis. DOE has also established a comprehensive plan for resuming production activities for the PT Facility that is similar to the process established and implemented for the HLW Facility. Resolution of the technical issues and establishment of a revised safety basis are key prerequisites to resuming engineering, procurements, and construction. (authors)

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DOE OSTI2019

Development of Polymer Waste Form for Spent Ion-Exchange Resins

A large amount of radioactive waste streams has been treated by ion-exchange resins at most of nuclear facilities for many years. These activities leave radionuclides as well as chemical contaminants on the spent resins, and the spent resins are considered as low- and intermediate-level radioactive waste. Spent ion-exchange resins require special immobilization treatment to meet the waste acceptance criteria for disposal. Immobilization of spent ion-exchange resins can be processed by mixing or incorporating them with suitable solidifying materials, such as cement, bitumen, organic polymers, and other materials or a combination of them. The selection of immobilization options for spent ion-exchange resins must consider their physical, chemical, and radiological characteristics. In this study, direct immobilization method was used to solidify the spent ion-exchange resins using organic polymer for improving the waste content. The objective of this study is to test the immobilization of spent ion-exchange resins using polymer waste form and enhance the solidified waste content. The specific objective of this study was to develop polymer waste form to solidify spent ion-exchange resins. Leaching (ANSI/ANS 16.1) and compressive strength tests were performed to meet the acceptance criteria of waste form for disposal. Ion-exchange resin has been solidified physically and chemically stable in the polymer waste form. The leaching test (ANSI/ANS16.1) for two polymer waste form specimens were conducted. The leachability indexes of sample A and B showed to meet the acceptance criteria for disposal. The compressive strength tested showed higher values in polymer waste form prepared with ion exchange resin compared to CoCl{sub 2} polymer form (sample B). Although spent resin was successively solidified in polymer matrix, the future experiment will develop a polymer waste form as a pre-treatment of spent resin for size reduction using a ball milling (500 RPM, 280 min)

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DOE OSTI2018

To the 'Safety related?' Question: What is Related to Safety? - 18430

Most, if not all, national programs for radioactive waste management (RWM) pledge their overall commitment to safety or - in the case of radioactive waste disposal - to long-term safety. Therefore, it may be somewhat surprising to find that the term 'safety' is hardly defined in these programs. The same holds for some of the core international guidance literature on the deep geological repository (DGR) 'safety case' concept. With respect to stakeholder concern over the safety of geological disposal, it seems, however, advisable to seek common ground in the understanding of the idea of 'safety'. In this paper, we review and discuss the most relevant international guidance literature's provision (or non-provision) of definitions of 'safety' in the context of radioactive waste disposal. Against this background, we develop a tentative, practical, glossary-style definition of 'safety' in the DGR context. As a seed of discussion, it may help to eventually expose possible mismatches in the base assumptions of safety experts and other stakeholders. (authors)

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DOE OSTI2020

The German Site Selection Process - Implementation and Challenges - 20528

Federal Company for Radioactive Waste Disposal (Bundesgesellschaft fuer Endlagerung mbH; hereinafter referred to as BGE) is a state-owned company established in September 2016 following the 'Act on the Organisational Restructuring in the Field of Radioactive Waste Storage'. 'Repository Site Selection Act' (hereinafter referred to as StandAG) defines the site selection procedure, aiming to ensure the best possible safety for storing high-level radioactive waste in Germany for the duration of one million years. The selection procedure is a participatory, transparent, learning and self-questioning process based on scientific expertise. Possible repository sites are sought all over Germany, without excluding or predetermining any potential site at the beginning. The procedure consists of three phases. After each phase, the legislative decides on the regions for further consideration in the next phase of the site selection procedure. The current first phase (Phase I) is divided into two steps. Objective of Step 1 is to determine subareas of favorable geological conditions based on available data. Step 2 of Phase I will lead to identification of specific regions that are considered for further surface-based geoscientific and geophysical exploration. Potential host rocks are rock salt, clay rock and crystalline rock. As part of the site selection procedure, an overview of the status of the planning and implementation of R and D activities is given. Topics of particular interest are, for example, the adaptation of the maximum temperature of 100 deg. C at the container surface, the impact of long-term climate variations as well as minimally invasive exploration methods. (authors)

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DOE OSTI2017

Special Analysis: 2016-001 Analysis of the Potential Under-Reporting of Am-241 Inventory for Nitrate Salt Waste at Area G

The 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.

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DOE OSTI2016

Progress of the Czech Republic Deep Geological Repository Program - 16255

The Czech Republic operates two NPPs with 6 pressurized water reactors. Czech state is fully responsible for the disposal of all radioactive waste including spent nuclear fuel and it provides the Radioactive Waste Repository Authority (SURAO) established in 1997. According to the RWM Policy a deep geological repository (DGR) considered to be the only feasible option for the final disposal of spent fuel. The State Policy set out important milestones for DGR development - the selection of candidate sites by 2020 and the selection of a final site by 2025. The final commissioning of a DGR is planned for 2065. The initial stages of DGR development involved the screening of geological conditions of the whole territory. Granites would provide the most suitable rock environment for DGR siting purposes. By the end of 2012 SURAO will have proposed 7 sites for the potential geological surveys and the first stage of site characterization work. Licenses for geological survey for this stage are in the force since October 2015. Main goal of RD and D programme is to support the siting programme for the selection of the candidate sites, to adapt a DGR design on suitable sites and to support the safety case. SURAO decided to enhance the extent of in-situ RD and D at a depth corresponding to that of the future DGR. In 2013 SURAO commenced the construction of the Bukov Underground Research Facility in the southern section of the Rozna uranium mine at a depth of 600 m below the surface in crystalline rock environment. (authors)

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