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DOE OSTI2019

Effective Access Monitoring at Geological Repositories

Access points at a deep, mined geological repository (GR) for the disposal of spent nuclear fuel (SNF) and other nuclear wastes present potential diversion paths for nuclear material. Because C/S measures are not likely to be used underground, access to a GR will require unprecedented reliance on C/S measures to maintain continuity of knowledge (CoK) on SNF buried underground. We develop a model GR based on common features of GR designs from national programs in order to develop and optimize C/S measures for GR access points that maximize confidence that CoK is maintained on SNF underground. Critical access points identified in this study are surface entrances to (1) the GR ramp (2) the excavation shaft, (3) the main elevator shaft, and (4) the ventilation shaft. The first three are considered critical detection points (DPs), whereas the fourth is considered a non-critical DP. The reason for the distinction is due to the different design capabilities of shaft components: the first three (ramp, excavation shaft, main elevator) are all capable of being used to move material from the underground to the surface, whereas the ventilation shaft is not. Such capabilities are verified during periodic design information verification (DIV) inspections.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Determination of Activity of Neodymium(III) in Molten Eutectic LiCl-KCl using Electrochemical Methods

The pyrochemical recycling of actinides is a promising option for closing the nuclear fuel cycle. The technology is particularly attractive for recycle of metallic and molten salt fuels. The Republic of Korea is also taking a serious look at the application of pyrochemical treatment of used oxide fuel from Pressurized Water Reactors (PWR). The science behind pyrochemical recycling is being actively studied at the various national labs and universities. The Rare Earth Drawdown process is proposed to be included in a commercial pyrochemical recycling facility in order to minimize the amount of waste generated from the processing that will require permanent disposal in a geological repository. This process involves selective extraction of rare earth elements from electro-refiner salt. The salt can then be further processed to selectively remove Group I-II active metals and eventually be recycled back to the electro-refiner. The preferred method of separation of the rare earths from the salt is using electrolysis due to its inherent selectivity and ease of operation. The purpose of this study is to evaluate the effects of the presence of cesium (Group I) on the electrochemical behavior of neodymium in a molten eutectic LiCl-KCl salt. The activity and by extension the activity coefficient of a species determines many of the key electrochemical properties such as apparent reduction potential. In this study, we have undertaken a series of experiments to experimentally determine the activity coefficient of the dissolved neodymium(III) ions as a function of concentration, both in the presence and absence of cesium. This was done using a first of its kind electrochemical cell. Previously, activity of dissolved species has been determined using electrochemical cells in which equilibrium potentials are measured for the Nd/Nd(III) versus a Ag/AgCl reference electrode. There are several issues associated with such an approach. These electrodes are associated with several sources of errors; indeterminate values of Standard Reduction Potentials of (E{sup θ}) for Ag and Nd in molten salts, unknown liquid junction potentials and potential drifts. These uncertainties contribute errors to the experimental measurements. For this study, we employed a Nd/LiCl-KCl-NdCl{sub 3(Saturatedelectrode)}. There are several advantages with using such an electrode. If the expanded Nernst Equation is written for the aforementioned electrochemical cell, all of the terms cancel out except the activity term. This enables us to directly experimentally measure the activity of the species in solution. When we use a saturated NdCl{sub 3} reference electrode, the activity that is measured is defined as the activity versus a 'Saturated Solid State'. To determine the activity coefficients, we need to determine the activity versus the 'Liquid Standard Sate'. To convert the activity from solid standard state to liquid standard state, we correct for it using the hypothetical Gibbs free energy of fusion for neodymium chloride. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Robotics on KM-IT platform - 19107

The Knowledge Management Information Tool (KM-IT) is a web-based system developed to maintain and preserve the Department of Energy's (DoE's) knowledge base. The system was developed by Florida International University's Applied Research Center (FIU-ARC) with the support of the D and D community including the DOE Office of Environmental Management, and with the collaboration and support of the DoE's Energy Facility Contractors Group (EFCOG) and the former ALARA Centers at Hanford and Savannah. The KM-IT system is a community driven system tailored to serve the technical issues faced by the workforce across the DOE Complex. This KM-IT web-based interactive system is operational and available for use at www.dndkm.org and is currently comprised of the following modules: Web Crawler, Hotline, Technology, Hanford's ALARA Center Reports, Specialist Directory, Lessons Learned, Best Practices, Video/Picture Library, Vendors and Industry News. The Web Crawler dynamically searches through the KM-IT repository as well as the web and displays search results based on the search criteria. The Hotline allows registered users to post questions/problems related to D and D and to receive solutions from a subject-matter specialist. The Technology module serves as repository of D and D technologies and contains technology/ demonstration fact sheets and vendor information. D and D community members can add Lessons Learned, Best Practices, and Pictures/Videos to the D and D repository, after a formal approval process. The Vendor module provides a directory of commercial vendors who provide D and D related technologies, supplies, and services. KM-IT makes an excellent use of the knowledge that exists within the D and D community by allowing D and D project managers around the DOE complex to collaborate by sharing innovative ideas, past experiences, and practices and by maintaining a directory of subject matter specialists. FIU is working on developing a new knowledge base using the existing KM-IT framework to provide focused information on robotics. FIU will develop the robotics technology and vendor modules to provide information related to robotics technologies and its application to many different areas including D and D, remote systems and waste processing. This robotics knowledge base will provide the ability to search through the information repository as well as display customized summary information including robotics technology descriptions, benefits, limitations, videos and pictures. The vendor module will provide information related to industry vendors for these technologies and include updated vendor contact information. The vendor module will also provide a search capability based on keywords as well as categories. The robotics knowledge base will publish commercially available technologies as well as the technologies that are still in the research and development phase with national laboratories and academic universities. The robotics KM-IT will be developed in phases to support all of the features of the KM-IT platform and will be published on www.rkmit.org. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

D and D of a Large Waste Treatment Plant in Germany - 18458

With its 40 years of nuclear experience, GNS, the world leading supplier of casks for spent fuel, ILW and HLW, also offers services for conditioning and management of all types of radioactive waste, especially for the German NPP. Since 1985, GNS treats and packages ILW and LLW from the operations and phase-out of German NPP in its Duisburg plant. To enable its permanent safe final disposal, solid wastes are compacted and dried before being shipped to an interim storage facility or - later - the Konrad repository. The total premises cover 20.000 m{sup 2} (215.000 sq ft), of which 9000 m{sup 2} (97.000 sq ft) is controlled area. The Duisburg Plant has three large industrial warehouses (each 25 m x 136 m or 82 ft x 446 ft) with a total capacity of 10.200 m{sup 2} or 110.000 sqft. For the treatment of open contaminated waste, six enclosures were built. These enclosures have a special ventilation system including a specific filtration plant. In one enclosure a high compaction plant was used, in another enclosures a drying facility for drums was in operation. During peak periods, more than 900 tonne of waste were treated annually. One enclosure was mainly used for maintenance, restoration and extension of mobile facilities, which were normally used for campaigns in NPP. Due to the decreasing amount of operational waste during the nuclear phase out in Germany, GNS has decided to close the Duisburg Plant. In the first half-year of 2017, the last waste packages were treated but around 80 20'-containers are still located on site. The D and D started in January 2017. The presentation will show the first experience of decommission processes of the first three enclosures. The following steps including the techniques used will be explained in detail: - Dismantling and removal of the high compaction facility; - Dismantling and removal of the drying facility; - Cleaning and wipe tests of the empty enclosures; - Removal of the stainless steel plates of the floor; - Investigations of the lower floor parts for contamination; - Removal of contaminated underground drainpipes; - Decontamination of the concrete floor; - Dismantling of the ventilation system in three enclosures; - Dismantling of the steel constructions. An important aim during the dismantling is to reduce the amount of nuclear waste and maximize the fraction material for free release. Next to the first three enclosures, in one of the enclosure IV - VI, a 150 tons scrap shear had been installed before the construction of the enclosure building. The scrap shear has to be dismantled and handled by heavy duty lifting equipment. Only afterwards is it possible to start the decommissioning of the last three enclosures. Our goal is to transfer this very large controlled area and the whole plant to the site owner, the Duisburg Port company, after 2 years and 5 months and doing so in a way so that the industrial warehouses can be used again for all possible types of industries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

GDSA Repository Systems Analysis Investigations in FY 2024

The Disposal Research and Development (R&D) Program of the US Department of Energy (DOE) office of Nuclear Energy (NE-8) Spent Fuel and Waste Science and Technology (SFWST) Campaign is to conduct R&D on disposal of spent nuclear fuel (SNF) and high-level waste (HLW). The goal of the Geologic Disposal Safety Assessment (GDSA) within this project is to develop a disposal system modeling and analysis capability that supports the integrated modeling of coupled processes controlling disposal system performance of deep geologic repositories, including uncertainty. This report describes specific activities in the Fiscal Year (FY) 2024 associated with the GDSA Repository Systems Analysis (RSA) work package. The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and repository system performance models in crystalline, argillite, salt, and unsaturated alluvium potential host-rock environments, and to simulate and analyze these generic repository concepts and models using GDSA Framework toolkit, and other tools as needed.

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

Reviewing Spent Nuclear Fuel Dry Cask Storage Programs to Identify Challenges for an Integrated Waste Management System - 17535

Under the Nuclear Waste Policy Act (NWPA), the U.S. Department of Energy (DOE) has responsibility for removing commercial spent nuclear fuel (SNF) from nuclear power plant sites and disposing of the SNF in a deep geologic repository. However, the methods used by commercial nuclear power plants for onsite storage of their SNF define the starting conditions for DOE efforts to develop a waste management system that includes packaging, transportation, storage, and disposal of commercial SNF. The NWPA (as amended) also charges the U.S. Nuclear Waste Technical Review Board (Board) with evaluating the technical and scientific validity of DOE activities related to the management and disposal of SNF and high-level radioactive wastes (HLW). DOE's efforts related to developing an integrated program for storing, transporting, and disposing of SNF and HLW were the focus of a public meeting held by the Board on August 24, 2016, in Washington, D.C. To prepare for the public meeting, the Board's staff compiled information on the dry cask storage systems in service at nuclear power plant sites in order to identify issues that would need to be addressed by DOE in planning to transport SNF away from the sites. This paper is based on a briefing presented to the Board members by the Board's staff before the public meeting. Post-meeting findings and recommendations of the Nuclear Waste Technical Review Board are documented in the Board letter to the Department of Energy dated December 8, 2016. A note about terminology: in this paper, the phrase 'dry cask storage system' is generic and applies to a complete system used for dry storage of commercial SNF.Within these storage systems, commercial SNF is sealed inside a container called either a cask or a canister depending upon the specific type of storage system used. This distinction is discussed in more detail in the body of this paper. When referring to the containers holding SNF in particular systems, this paper uses the terms 'casks' or 'canisters,' as appropriate. When discussing the collective group of all containers, this paper uses the phrase 'casks and canisters'. Dry cask storage of SNF is a widespread practice throughout the U.S. commercial nuclear industry. As of August 1, 2016, more than 2300 storage casks and canisters were loaded at U.S. commercial nuclear power plant sites. This number is expected to increase by more than 200 casks and canisters annually, reaching a total of approximately 4500 by 2025. Multiple vendors have developed their own unique designs of dry cask storage and transportation systems. The majority of storage system designs (75 percent) feature SNF loaded in welded stainless steel canisters that are stored in ventilated concrete overpacks that allow for convection cooling. Some older dry cask storage system designs do not include a welded stainless steel canister; instead, SNF is loaded directly into a heavily shielded metal cask to minimize radiation exposure to workers and prevent atmospheric release of radionuclides. These types of systems are sealed by O-rings compressed under a bolted lid. After loading, dry cask storage systems are located on concrete pads at Independent Spent Fuel Storage Installations (ISFSI) at nuclear power plant sites and will remain there until the time when they will be transported to a consolidated interim storage facility or a deep geologic repository. Eventual transportation of canistered SNF will include removing the stainless steel canisters from the concrete overpacks, possibly repackaging the SNF, and loading the canisters into specially designed transportation casks for shipment by truck, rail, or barge. In the case of the non-canistered casks, some are approved for transportation and can be shipped after impact limiters are installed on the casks. SNF will need to be removed from casks not approved for transportation and transferred to transport-approved casks. This paper highlights five issues and challenges presented by the nuclear industry's use of dry cask storage systems that will need to be addressed by DOE in developing an integrated waste management system. One example, among the five, is that several dry-storage casks and canisters have received Nuclear Regulatory Commission certificates of compliance for storage only and do not have certificates of compliance for transportation. This will pose a challenge when the SNF must be moved to a consolidated interim storage facility or a deep geologic repository. Another example is that, when loaded with SNF, the total weight of new, larger transportation cask designs will exceed 142 metric tonnes (156 tons) which, when combined with the weight of the rail-car, will exceed the gross weight limit for unrestricted operation on some rail lines. This will also affect future transportation planning for shipments of SNF. All five issues and challenges are discussed in detail in the body of the paper. (authors)

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

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cask storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather, the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI) which has only been demonstrated on a limited basis, a highly oxidizing system is required. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cast storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI), a highly oxidizing system is required, which has only been demonstrated on a limited basis. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Cigeo, The French Geological Disposal Facility Project: Preliminary Design (APS) Phase - 16038

The Cigeo project has been in the works for 25 years. Numerous studies have been conducted, increasingly associated with the Meuse/Haute-Marne site, with further specific research thanks to direct access to the Callovo-Oxfordian clay formation from the underground laboratory of Bure-Saudron. These studies and research initially aimed to demonstrate the feasibility of the repository. They also helped gain a high level of understanding of phenomena to support design studies and demonstrate safety. Transition to the industrial phase began with the development of a plan for delivering waste to the facility for disposal. The plan introduced sequencing for the various types of waste to be disposed of, and was optimised to determine the size of inspection, transfer and handling facilities. In describing the life of the repository and therefore the vision for its operation, it has become obvious that our generation should not impose choices on future generations. We must provide them with reference technical solutions, with the financial resources to implement them. It is also our duty to begin the construction and initial operating phases. However, because the facility will operate over several generations, we must leave a degree of flexibility so that they may reassess the options that we define and adopt their own solutions, as necessary. They will also benefit from operational experience gathered as facility operations develop. This is the context in which the preliminary design phase is being finalised in preparation for the detailed design phase, with the aim of gradual commissioning during the latter part of the next decade. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Near Repository Unsaturated Alluvium Disposal Modeling with Improved Geological Realism

The Disposal Research and Development (R&D) Program of the US Department of Energy (DOE) office of Nuclear Energy (NE-8) Spent Fuel and Waste Science and Technology (SFWST) Campaign is to conduct R&D on disposal of spent nuclear fuel (SNF) and high-level waste (HLW). The goal of the Geologic Disposal Safety Assessment (GDSA) within this project is to develop a disposal system modeling and analysis capability that supports the integrated modeling of coupled processes controlling disposal system performance of deep geologic repositories, including uncertainty. This report describes a specific activity in the Fiscal Year 2024 (FY24) associated with the GDSA Repository Systems Analysis (RSA) work package in collaboration with the GDSA Geologic Modeling work package at Los Alamos National Laboratory (LANL). The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and repository system performance models in crystalline, argillite, salt, and unsaturated alluvium potential host-rock environments, and to simulate and analyze these generic repository concepts and models using GDSA Framework toolkit, and other tools as needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Cost Implications of an Interim Storage Facility in the Waste Management System

This report provides an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). Specifically, the impacts of the timing of opening an ISF relative to opening a repository were analyzed to understand the potential effects on total system costs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

GDSA framework, a computational framework for complex modeling problems in radioactive waste management

This paper details a computational framework to produce automated, graphical workflows, and how this framework can be deployed to support complex modeling problems like those in nuclear engineering. Key benefits of the framework include: automating previously manual workflows; intuitive construction and communication of workflows through a graphical interface; and automated file transfer and handling for workflows deployed across heterogeneous computing resources. This paper demonstrates the framework's application to probabilistic post-closure performance assessment of systems for deep geologic disposal of nuclear waste. However, the framework is a general capability that can help users running a variety of computational studies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Experiments and Modeling to Support Field Test Design

Salt repositories continue to be a primary concept for disposal of heat-generating nuclear waste (HGNW). For the past several years, efforts have focused on the in-drift disposal concept (e.g. Hansen and Leigh 2011; Stauffer et al. 2013; Jordan et al. 2015a,b,c; Bourret et al. 2016; Hansen et al. 2016; Bourret et al. 2017; Johnson et al. 2017a,b; Kuhlman et al. 2017). Ongoing research efforts aim to improve understanding of the behavior of the salt backfill surrounding HGNW sources, the responses of drift walls to heat sources, and the chemistry and chemical evolution of brine in such environments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Corrosion of U233-Doped Uranium Oxide using Microfluidics Methods

The aim of work this year has been to investigate the role of alpha (α)-radiation and hydrogen (H 2 ) on the corrosion of uranium oxide (UO 2 ) using a microfluidic device. The microfluidic device, termed the Particle-Attached Microfluidic Electrochemical Cell, (PAMEC), enables monitoring of the UO 2 electrochemical corrosion potential (E corr ) under different environments, including de-aerated conditions and in the presence of dissolved H 2 . The Si 3 N 4 window allows us to study morphological and chemical changes under an electron microscope. We have previously demonstrated that the PAMEC matches the results from bulk electrochemical tests with UO 2 [1]. We used the high specific activity uranium (U) isotope, 233 U, (t 1/2 = 160,000 years) incorporated into UO 2 , to generate a localized a-field. The objective of the experiments were to mimic the radiation environment that would be experience at the surface of aged spent nuclear fuel (SNF) during long-term geologic disposal under anoxic conditions. Wittman et al. [2] predicted that in the presence of a pure a-radiation field and under H 2 conditions, the concentration of the radiolytic oxidant H 2 O 2 would be suppressed or even eliminated. In a UO 2 corrosion experiment this would be exhibited through a lowering of the measured UO 2 corrosion potential compared to identical conditions in the absence of dissolved H 2 . We found that the predictions of Wittman and co-workers were supported and that the corrosion potential of the 233 U-doped UO 2 in solution lowered with presence of H 2 gas and increased in the absence of H 2 , when under anoxic conditions. The PAMEC experiments indicate that H 2 O 2 has been eliminated in a solution sparged with H 2 while exposed to an a-radiation field. The corrosion potential of the a-doped 233 U(10%)- 238 UO 2 in a solution sparged with Ar/H 2 matched the corrosion potential of 238 UO 2 in solution sparged with air. This clearly demonstrated that the H 2 O 2 had been eliminated and that the only oxidant present was O 2 in this system in complete agreement with the modeling results of Wittman et al. These results point to the need to improve the Fuel Matrix Degradation (FMD) Process Model training data set that is being used in the FMD surrogate model that is being developed for the repository program. The incorporation of realistic radiation chemistry will improve the scientific basis for the FMD Model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Physical characterization of fault rocks within the Opalinus Clay formation

Near-surface disposal of radioactive waste in shales is a promising option to safeguard the population and environment. However, natural faults intersecting these geological formations can potentially affect the long-term isolation of the repositories. This paper characterizes the physical properties and mineralogy of the internal fault core structure intersecting the Opalinus Clay formation, a host rock under investigation for nuclear waste storage at the Mont Terri Laboratory (Switzerland). We have performed porosity, density, microstructural and mineralogical measurements in different sections of the fault, including intact clays, scaly clays and fault gouge. Mercury intrusion porosimetry analysis reveal a gouge that has a pore network dominated by nanopores of less than 10 nm, yet a high-porosity (21%) and low grain density (2.62 g/cm 3 ) when compared to the intact rock (14.2%, and 2.69 g/cm 3 ). Thus, a more permeable internal fault core structure with respect to the surrounding rock is deduced. Further, we describe the OPA fault gouge as a discrete fault structure having the potential to act as a preferential, yet narrow, and localized channel for fluid-flow if compared to the surrounding rock. Since the fault gouge is limited to a millimetres-thick structure, we expect the barrier property of the geological formation is almost not affected.

58 GEOSCIENCES↗
DOE OSTI2020

The Impact of Ethylenediaminetetraacetic Acid (EDTA) on the Sorption of Nd(III), Th{sup (IV)}, and U(VI) onto Dolomite in WIPP-Relevant Brines, GWB and ERDA-6 - 20356

Storage of legacy, transuranic waste at the Waste Isolation Pilot Plant (WIPP) requires accurate risk models for performance assessments. These models must predict the long-term fate and transport of contaminants in the WIPP and surrounding environment. Laboratory support is necessary for the development of an accurate model through better understanding of contaminant behavior in the WIPP. The WIPP is a deep geologic repository that is characterized by high ionic strength due to evaporite deposits in the Salado formation (porewaters up to 7.4 M). High concentrations of sodium chloride (NaCl), calcium chloride (CaCl{sub 2}), and magnesium chloride (MgCl{sub 2}) are particularly of concern due to their potential impact on solubility and ultimately transport of contaminants of concern. Additionally, ligands capable of forming strong complexes with metals are found in significant concentrations in the WIPP environment due to their disposal from waste processing and formation during degradation of repository components (e.g. cellulose degradation by calcium hydroxide in cement). Consequently, there is a need to investigate the fate and transport of contaminants in the WIPP environment with a focus on the effects of high ionic strength and strongly coordinating ligands. Contaminants of high concern include the actinides (U(VI), Pu(IV), and Am(III)). Laboratory support of the WIPP modeling efforts was provided by this research. A series of batch experiments were utilized to study the impact of ionic strength, contaminant concentration, and the presence of ethylenediaminetetraacetic acid (EDTA) on the sorption of Nd(III), Th(IV), and U(VI) onto dolomite, a carbonate mineral found within the Culebra formation. The Culebra formation is the most transmissive layer above the WIPP, making transport of contaminants the most likely route in previous risk assessments. Two brines specifically designed to imitate conditions within the WIPP environment were used in these experiments. The U.S. Energy Research and Development Administration Well 6 (ERDA-6, 95%) brine simulates a low Mg environment, while the generic weep brine (GWB, 95%) simulates the WIPP environment with a high Mg concentration. Mg is added to the repository in the form of MgO as an engineered barrier and is naturally present in some formations. It is expected to scavenge carbonate from solutions decreasing formation of the highly soluble and competitive uranium-carbonate complexes. EDTA is used during manipulations of nuclear materials due to its ability to form strong, stable complexes with the actinides and may be found in high concentrations (up to 0.08 mM) within the WIPP as waste canisters degrade. Further, it may provide additional avenues for unwanted transport of contaminants. Nd(III) and Th(IV) were used as stable analogues of Am(III) and Pu(IV), respectively, to represent the most common oxidation states of actinides in the WIPP environment in addition to U(VI). The potential for colloid formation was also investigated with low and high concentrations of contaminants (10 and 1000 μg/L) using three size-dependent particle separation steps. The results of this research provide a better understanding of potential behavior of actinides in the presence of EDTA and within WIPP-relevant brines. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Integrating Management of Spent Nuclear Fuel from Generation to Disposal

The current management system in the United States for commercial spent nuclear fuel does not emphasize integration among storage, transportation, and disposal. Unless a path can be implemented that addresses the long-term needs for integration, the United States could end up leaving substantial quantities of stranded commercial spent nuclear fuel stored at decommissioned reactor sites in an increasingly wide variety of containers. This lack of integration does not cause safety issues, but may delay transporting the spent fuel and complicate options for permanent disposal. The large containers now in use for dry storage remain at high temperatures for decades, thereby delaying transportation from decommissioned reactors. The large containers also have no easy path to disposal unless (1) disposal is further delayed (up to 150 years or more for some mined repository concepts); (2) the contents are repackaged into smaller, cooler packages; or (3) the high temperatures are used as de facto site-selection and design criteria for a repository. Implementing consolidated interim storage could address many issues that exist because of this lack of integration. A consolidated interim storage facility that includes appropriate capabilities can allow existing disparate parts to integrate as a system. Previous agencies and commissions have noted this theme before as a way to provide flexibility in the waste management system. This report uses the rationale for such an approach as a framework to discuss the complexities of reconfiguring the waste management system to include consolidated storage. However, concerns that increased storage capacity will reduce the national urgency for a repository are unavoidable, and continued effort will be necessary in public dialogues on the societal aspects of moving commercial spent nuclear fuel into consolidated interim storage. A single optimal solution for integrating current storage and planned transportation with disposal is unlikely. Rather, efforts to integrate various phases of spent fuel management should begin promptly and continue throughout the remaining life of the current fuel cycle. These efforts will need to adapt continuously to evolving circumstances.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

RH-TRU Waste Streams: HFDA Crucibles Create New Complex Waste Stream

This project focuses on the removal of the Hot Fuel Dissolution Apparatus (HFDA) crucibles from the Hot Fuel Examination Facility’s (HFEF) hot cells to create space for upcoming projects. The project team has created a crucible and pyro-salt non-debris waste stream that can be disposed of at Waste Isolation Pilot Plant (WIPP). Chemical Compatibility Examinations (CCEs) must be performed on candidate waste by reviewing the types of chemicals used in the experiments, information found in INL Process Knowledge Summary Reports (PKSRs), and Environmental Protection Agency guides on chemical compatibility. The CCEs identify which chemicals can be safely combined to ensure adverse reactions do not occur in the waste packages. The CCEs are then used to update the PKSR, Basis of Knowledge, and Acceptable Knowledge documents and demonstrate compliance with the WIPP Waste Acceptance Criteria. Once documents are finalized, they are reviewed and approved by the WIPP Central Characterization Project (CCP) to ensure proper waste stream characterization. After the documents are approved, the crucibles can be packaged and shipped to the Idaho Nuclear Technical and Engineering Center for storage awaiting certification. CCP will certify the waste using the approved documents and a series of confirmatory testing before being shipped and disposed in WIPP’s deep geological repository. Currently, the CCE is in progress and the PKSRs are being updated.

Basis of Knowledge↗