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Used Nuclear Fuel Disposition R&D Campaign and Waste Form Impacts on Repository Performance.
Abstract not provided.
Debrief for the DOE-Managed Spent Nuclear Fuel and High Level Waste Research (aka Defense Repository).
Abstract not provided.
Introduction to the DOE-Managed Spent Nuclear Fuel and High Level Waste Research (aka Defense Repository).
Abstract not provided.
Enhanced Performance Assessment Models for Generic Deep Geologic Repositories for High-Level Waste and Spent Nuclear Fuel - 16223
Development of an enhanced performance assessment (PA) capability for geologic disposal of spent nuclear fuel and high-level waste has been ongoing for several years in the U.S. repository program. The new Generic Disposal System Analysis (GDSA) modeling and software framework is intended to be flexible enough to evolve through the various phases of repository activities, beginning with generic PA activities in the current Concept Evaluation phase to site-specific PA modeling in the Repository Development phase. The GDSA Framework utilizes modern software and hardware capabilities by being based on open-source software architecture and being configured to run in a massively parallel, high-performance computing (HPC) environment. It consists of two main components, the open-source Dakota uncertainty sampling and analysis software and the PFLOTRAN reactive multi-phase flow and transport simulator. Reference cases or 'generic repositories' have been, and are being developed, based on typical properties for potential salt, clay, and granite host-rock formations and corresponding engineered design concepts for each medium. Past simulations have focused on a generic repository in bedded-salt host rock, while the most recent research has focused on a reference case for a typical clay/shale host rock. A variety of single-realization (i.e., deterministic) and multi-realization (probabilistic) results for the new clay reference case are presented, including an analysis of the effects of heat generation on repository performance, assuming a 100-year out-of-reactor commercial SNF waste form. Order-of-magnitude differences between predicted radionuclide concentrations in thermal versus isothermal simulations imply that mechanistic, coupled-process modeling in three-dimensional (3-D) domains can be important for building confidence in post-closure performance assessments. (authors)
What's in WIPP? Packaging TRU Waste to Enhance WIPP's Capacity - 16515
Beginning in 1999, and continuing for 15 years, the US DOE Waste Isolation Pilot Plant (WIPP) repository successfully disposed of defense Transuranic (TRU) waste. Six of the originally planned ten disposal panels were filled before two unrelated accidents temporarily halted disposal operations in February 2014. Based on counting the gross volume of all containers emplaced in WIPP to date, a little more than half of the legislated volume capacity has been subscribed (91,000 cubic meters). This paper describes the many different packaging configurations that have been used to ship TRU waste to WIPP, and makes observations that may enhance the repository capacity when shipments resume. The contents of shipments to WIPP are limited by many requirements of the repository waste acceptance criteria. These criteria derive from three primary sources; 1) the Certificate of Compliance (CoC), issued by the NRC for Type B shipping packaging used by WIPP; 2) The Hazardous Waste Facility Permit, issued by the New Mexico Environment Department; and 3) the WIPP facility's own Documented Safety Analysis (DSA), developed under DOE's regulation 10CFR830. To meet these criteria, various TRU waste forms have been packaged in different ways, with varying packaging efficiency. Of course, there is an overall upper limit on WIPP's volume capacity that was legislated as part of the WIPP Land Withdrawal Act. The volume counted against this limit has been calculated by simply summing the volumes of individual payload containers shipped to and emplaced at WIPP. This method counts significant non-waste volumes inside the containers due to over-packing inner containers and partially filled containers. Without some change to how disposal volumes are calculated, or how waste is packaged, the amount of TRU waste yet to be shipped to WIPP could exceed the legislated capacity. Concepts to enhance the packaging efficiency of future waste shipments to WIPP are presented. Approximate estimates of the waste versus non-waste volumes that have been emplaced to date are discussed. This paper only considers contact handled TRU waste (i.e., contact dose rates < 2 mSv/hr). (authors)
Influencing future exploratory drilling rates-a potential approach for the waste isolation pilot plant - 15307
The Department of Energy/Carlsbad Field Office (DOE/CBFO) is responsible for managing activities related to the disposal of TRU and TRU-mixed waste in the geologic repository at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The WIPP Compliance Certification Application had a reference concerning natural resources at the site of the Waste Isolation Pilot Plant. This reference suggested that resources could be withdrawn from under the repository area from a distance and not affect the integrity of the repository. This would not eliminate, but would certainly diminish, the rate of exploratory drilling in the distant future. Currently the only scenarios that result in releases from the repository are human intrusion scenarios, without human intrusion radioactive releases to the accessible environment are essentially zero for the very long term future. (authors)
Resolving Issues Regarding Disposal of Safeguarded Nuclear Materials: Concepts for Near Surface and Intermediate Depth Disposal sites
This document presents two case studies on nuclear waste disposal to explore the safeguards challenges and potential solutions present under each scenario. Both involve the disposal of nuclear materials, either for waste containing safeguarded nuclear materials or waste where safeguards on the nuclear materials have been previously terminated. The first case study is that of a low-level waste (LLW) repository that is situated near the surface. This facility will accept safeguarded nuclear materials for disposal. Examples of this type of activity are rare, but the potential for more facilities of this type is very possible as States become more comfortable with considering safeguards measures in perpetuity on deep geological repositories. The second study extends this to the case of terminated wastes at an intermediate depth disposal site. In this case, the reasons why terminated nuclear materials could be found in these disposal facilities and how obligations under the Comprehensive Safeguards Agreement (CSA) and Additional Protocol (AP) are met when the disposed waste items are intermediate level waste (ILW) or high-level waste (HLW).
Communicating with future generations - Major obstacles to overcome - 15009
The Department of Energy (DOE) Carlsbad Field Office (CBFO) is responsible for managing activities related to the disposal of transuranic (TRU) and TRU-mixed waste in the geologic repository at the Waste Isolation Pilot Plant (WIPP), 2150 feet (655 meters) below the land surface, at the WIPP, near Carlsbad, New Mexico. The Environmental Protection Agency (EPA) regulations Title 40 Code of Federal Regulations Part 191.14(c) [1] and Title 40 Code of Federal Regulations Part 194.43 [2] require a Passive Institutional Controls program be developed for the WIPP. The primary purpose of the Passive Institutional Controls program is to indicate the location of the repository and the risk it presents to an intruder, thus reducing the likelihood of inadvertent human intrusion into the repository. A plan was put into place during the 1990's with records management and storage, awareness triggers, permanent marker design concepts, including languages, graphics and engineered drawings, plus testing schedules for materials and structures. This work included the thoughts of expert panels and individuals. The plan held up under peer review and met the regulatory requirements. But will it really work? Will future generations a thousand years, two thousand years, or seven thousand years from now really be able to decipher and understand the message? This is the challenge international constructors, operators, and regulators of repositories for the disposal of any hazardous materials must face. It is important to explore the history of language, both written words and graphics; and what we can learn from the past to assist us in communicating with the future. (authors)
Pseudo-dynamic Aspen Plus{sup R} Model of the Defense Waste Processing Facility - 18475
The Defense Waste Processing Facility (DWPF) at the Savannah River Site converts legacy high-level waste from the Cold War into borosilicate glass for permanent disposal in a geological repository. The waste sludge is chemically adjusted by the addition of acid, mixed with frit, and concentrated in the DWPF Chemical Process Cell (CPC) to ensure that the final glass has the desired properties. The resulting mixture is fed into the melter where the waste is melted and poured into stainless steel storage canisters. We present an Aspen Plus{sup R} model of the DWPF CPC, which is being developed to aid chemical processing decisions pertaining to the current flowsheet and to aid decisions regarding optimizing future processing. Given a waste feed composition with acid and frit additions, the model will predict the sludge and off-gas compositions throughout the processing, as well as the final melter feed composition. This allows different processing scenarios to be investigated easily, reducing the reliance on experimental testing, and hence saving time and money. The model has multiple potential applications, including calculating the acid and frit additions required to give acceptable glass properties, off-gas permitting, tracking mercury steam-stripped from the waste in the off-gas system, tracking flammable gases in the offgas system, investigating the effect of reducing air purges in the condensers, and general process optimization. The waste is processed in the CPC in batches, requiring dynamic simulation to track the time-dependent operations. Accurate modeling of the processing also requires calculations of phase and electrolyte chemical equilibria and parallel rate-dependent reactions. Due to existing technical limitations with dynamic simulation of the complex chemical system, we have adopted a pseudo-dynamic approach in which time intervals are treated as consecutive steady-state operations in Aspen Plus{sup R}. Thus, the model consists of ∼100 replicas of the processing vessel and its off-gas system, with the sludge output by one replica (representing one time interval) becoming the feed to the next (representing the next time interval). By examining the results from these replicas in sequence, one can see the time evolution of the model. This general approach can be used wherever simultaneous calculation of phase and electrolyte chemical equilibria in a dynamically evolving system is needed. In addition to describing the general structure and chemistry of the model, we present results from tests of the model against experimental data. These data are from laboratory-scale process simulations of the CPC with a nonradioactive waste sludge simulant. These tests include the model's predictions of the offgas composition, the sludge pH, and the final sludge composition. We will also discuss planned future developments of the model. (authors)
Insights into the thermo-hydraulic properties of compacted MX80 bentonite during hydration under elevated temperature
In high-level radioactive waste geological repositories, compacted bentonite undergoes coupled thermo-hydraulic processes due to heat released from a central waste canister and groundwater imbibition from the surrounding host rock. An understanding of these processes is essential for long-term simulations of radionuclide migration and canister corrosion, which requires an understanding of temperature effects on the coupled thermo-hydraulic properties governing these processes. In this study, a tank-scale radial infiltration test was used to investigate water imbibition processes in compacted bentonite under a central heater temperature of 200 °C that simulates high thermal gradients in a repository. Interpretation of this test focuses on evaluation of the liquid water wetting front during hydration and the interpretation of the temperature-dependent transient soil water retention curve (SWRC), thermal conductivity function (TCF), and hydraulic conductivity function (HCF). The SWRC during imbibition follows a temperature-dependent wetting path. Temperature effects on the hydraulic conductivity of bentonite in saturated conditions had the greatest effect on the shape of the HCF, with minimal temperature effects at higher suctions. The transient thermal conductivity data matched well with a new TCF linked with the SWRC shape. Although the bentonite layer was restrained, local deformations during hydration may have affected the shapes of the TCF and HCF.
Performance Assessment of a Generic Repository for Defense-Related HLW/SNF in Fractured Crystalline Host Rock - 17059
Recently the U.S. made a policy decision to begin R and D for geologic disposal of defense-related HLW/SNF in a facility separate from commercially generated waste. A deep geologic repository for the disposal of DOE-managed HLW, and some thermally cooler DOE-managed SNF, arising from defense and DOE R and D activities, is part of a stepwise, phased approach for disposal of the Nation's nuclear waste, as recommended by the Blue Ribbon Commission on America's Nuclear Future. The work discussed here focuses on post-closure safety (or performance) assessment for such a Defense Waste Repository (DWR), which is divided into four major activities: (1) development of generic reference cases (i.e., knowledge or technical bases for 'generic' or 'non-site-specific' deep geologic repositories); (2) features, events, and processes (FEPs) analyses and screening to support the technical bases and the performance assessment (PA) model; (3) performance evaluation of alternative EBS design concepts; and (4) post-closure safety analyses of the repository system under consideration. Using the known inventory of defense-related SNF, as well as defense-related HLW stored at the Savannah River and Hanford sites, the Geologic Disposal Safety Assessment (GDSA) Framework modeling and software system has been applied to simulate the potential performance of a DWR in crystalline host rock, resulting in a suite of single-realization (i.e., deterministic) and multi-realization (i.e., probabilistic) 3-D post-closure system analyses, over a performance period of one million years. Two types of emplacement concepts are examined, including single-canister vertical-borehole emplacement for the hotter defense SNF waste (KBS-3V concept) and multi-canister horizontal emplacement for defense HLW (similar to Yucca Mountain co-disposal waste packages). Sensitivity analyses examine the effect of key uncertain parameters on repository performance, including the effects of fracture distribution, waste package degradation rate, buffer and disturbed rock zone (DRZ) properties, and sorption parameters. Initial results indicate that a crystalline host rock with a connected fracture system may require additional safety features to ensure robustness of the isolation safety function, such as a deep unsaturated zone, a sufficiently thick sedimentary overburden, and/or a disposal overpack with a very slow corrosion rate. None of these should pose an undue obstacle for successful disposal. (authors)
Uranium(VI) Sorption onto Montmorillonite Clay in the Presence of Ca2+/Calcite Impurities
The storage of nuclear waste creates a series of significant environmental risks. One attempt to minimize these potential impacts is the use of underground geological repositories, where waste containers are surrounded by geologic buffer materials creating an engineered barrier. One buffer material that has been of interest is bentonite, which is largely composed of montmorillonite clay. These barriers limit the mobility of radioactive contaminants, such as uranium(VI), into environmental systems based on contaminant sorption onto the clay material. System conditions, such as pH, the presence of mineral impurities such as calcite, and the concentrations of U(VI) and atmospheric CO2 also affect the sorption behavior of U(VI) onto these clay barriers. This study focuses on the effects of Ca2+ and calcite (CaCO3), a common impurity found in montmorillonite on the sorption behavior of U(VI). Computer simulations of the chemical solution species (chemical form) of U(VI), in the absence and presence of Ca2+/calcite impurities are performed in parallel with laboratory sorption experiments. Experimental data will be compared to modeling results to help obtain further insights regarding the impacts of Ca2+/calcite impurities on U(VI) solution speciation and sorption behavior onto montmorillonite.
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)
An Overview of R and D on Retrievability and Retrieval Technology in Germany - 20462
Retrievability is a term that is included in most radioactive waste management programs around the world. Although national definitions vary, the overall understanding of retrievability concerns the ability to recover waste packages from the repository mine after their emplacement. Different countries may implement retrievability in very different ways, ranging from a built-in reversibility into the emplacement process to the stipulation that retrievability may in no way impede passive safety in the post-closure phase. Germany takes a middle course such that retrievability in a HLW repository (repository for high level radioactive waste) may have no significant detrimental effect on passive safety. In Germany, the current siting process considers rock salt, clay rock, and crystalline rock as potential host rocks. Therefore, Research and Development (R and D) has been investigating repository concepts and retrievability in all host rocks. After introduction of retrievability in 2010, existing repository concepts were modified to facilitate retrieval. The changes made comprised, for example, equipping boreholes with steel liners, developing new technologies, and modifying existing emplacement devices. Apart from retrieval of HLW, retrieval of other wastes from underground repositories that were designed and operated without retrievability in mind, poses major technical and scientific challenges. Currently, studies are under way to investigate the feasibility and costs of partial retrieval of waste from an underground repository for hazardous and highly toxic waste in France, Stocamine. With regard to retrieval, the rock-mechanical conditions are deteriorating rapidly, so time is of the essence. In the Asse II mine in Germany, about 126,500 waste casks with low and intermediate level waste await retrieval. The repository mine suffers from difficult rock-mechanical conditions and inflow of brine, locally into emplacement areas. The Federal Company for Radioactive Waste Disposal (BGE) is legally required to retrieve all wastes from the Asse. Due to the specific challenges, R and D is needed to develop technical solutions for the safe retrieval from each of the emplacement chambers. R and D on retrievability and retrieval technology does not only address challenges in radioactive or toxic waste retrieval but may also help to germinate innovation to better master complex underground situations in general, e.g. in deep mining, tunneling, or repository construction. (authors)
Water Migration and Swelling in Engineered Barrier Materials for Radioactive Waste Disposal
Deep, underground repositories are needed to isolate radioactive waste from the biosphere. Because bentonite is an integral component of many multibarrier repository systems, information on the hydraulic behavior of bentonite is crucial for modeling the long-term viability of such systems. In this paper the hydraulic behavior of bentonite samples was analyzed as a function of aggregate size, and samples were subjected to hydrothermal treatments involving contact with NaCl, KCl, and deionized water. Neutron and X-ray imaging were used to quantify water sorption into packed bentonite samples and bentonite swelling into the water column. The distance between the original clay-water interface and the wetting front was determined as a function of time. Average water uptake exhibited a square-root-of-time dependence in freshly prepared samples, but more variable rates were observed for samples previously in contact with water. The radiography was supported by small-angle neutron scattering analysis and ultra-small-angle neutron scattering analysis of aggregate size distributions and by inelastic neutron scattering to understand the physicochemical environment of the sorbed water. Results showed that hydrothermal treatment with KCl had the greatest effect of increased water transport in the bentonite, possibly as a result of the interaction of K + with smectite layers in the clay.
Management of Metals from Decommissioning of Nuclear Facilities - 18411
The EDF group companies within the Cyclife platform have for decades performed treatment and recycling of metals. With different approaches but all considered international Good Practice, as valuable materials are recovered for safe recycling and volumes in the repositories for radioactive waste are saved. For an efficient decommissioning project, a well-defined waste management process is one of the critical paths, not least when it comes to routes for clearance and recycling of materials. It is easy to underestimate the complexity and bottlenecks caused by shortage of logistic space. In order to mitigate these risks, activities such as strengthening categorization, quantification and adopting different routes based on the material properties should be implemented. By categorization of the of waste and materials considered/classified as contaminated or potentially contaminated it is obvious that the management of the metals is a key parameter in nuclear decommissioning. Analyses performed by the EDF companies shows that a majority of the materials, classified as contaminated, from a BWR/PWR plant are metals. Experience gives that a high percentage of these metals from decommissioning can be subject to recycling either directly upon completion of the clearance process or after certain treatment. Non-technical key parameters like the regulatory requirements on applying the waste hierarchy, the impact on repository programs as well as, not at least, the regulatory and public acceptance of recycling of materials from the nuclear industry are, together with costs essential factors when forming the routing of metals and other waste materials from a decommissioning project. Today the EDF group companies are positioned to provide a wide range of services for management of the contaminated metals from decommissioning projects, spanning from industrial management of metals which directly will be subject general clearance after verification measurements up to decontamination and melting of selected reactor internals for recycling to public domain or within the nuclear industry. In this paper the typical waste material composition for BWR and PWR plants, the different ways to treat/manage the radioactive and potentially radioactive metals including a comparison of the different approach commonly used in a technical, logistical and economical perspectives. A few success stories will be resented as well as lessons learned. (authors)
Transportation of Low-Level Radioactive Waste from KAERI to the Final Repository - 18179
The Korea Atomic Energy Research Institute (KAERI) collects the radioactive wastes generated during the operation of nuclear facilities and R and D processes. Radioactive wastes are treated and repackaged according to the national radioactive waste management policy, and then transferred to the final disposal site at the Korea Radioactive Waste Agency (KORAD). After 2015, the wastes delivered to the final disposal site have consisted of spent filters and combustible wastes such as clothes and gloves generated during R and D or facility operation processes at KAERI. The wastes generated by each facility are collected at the radioactive waste treatment facility, then treated and repackaged in 200 L drums according to their generation characteristics, i.e. each drum contains waste of the same type from the same facility, and with the same date of generation. Then, a radionuclide inventory is obtained for the repackaged drum by using a direct measurement method such as the drum assay system, radiochemical analysis through random sampling or an indirect measurement method such as applying a scaling factor. When performing sampling of the radioactive waste drums, the waste samples with a high contamination are selected in a conservative approach as representative. The radioactive waste drums are inspected and checked for acceptability by the Quality Management Division of KAERI, and a preliminary inspection is carried out on-site by KORAD. At this time, the acceptability of the radioactive waste drums is assessed through visual inspection, the appropriateness of their nuclide analysis value is confirmed, and a survey or smear test of sample drums is performed. Drums that meet the acceptance criteria are transferred to KORAD after a radiation environmental impact assessment of the transportation. Since 2015, 1,084 low-level operational radioactive wastes have been transferred from KAERI to the final disposal site, and further preparation work is now being made for the transportation of additional drums. (authors)