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GDSA - Modeling and Integration – ORNL: PFLOTRAN modeling

This report summarizes work done at Oak Ridge National Laboratory (ORNL) related to post-closure criticality consequence analysis of dual-purpose canister (DPC) direct disposal in unsaturated alluvium. PFLOTRAN software was used to model the behavior of a breached DPC with low water infiltration rates and unsaturated conditions in alluvial deposits such that a low-power criticality event occurs with a focus on thermal hydrological constraints on criticality timing and power output. This summary describes initial exploration of those thermal hydrological processes and of the long-term average power output from a criticality event that could be sustained without driving water out of the DPC package and thereby terminating the criticality event. The objectives are (1) to identify, for different groundwater infiltration rates, the time at which the DPC fills with water sufficiently to potentially initiate criticality; (2) to bound the power output that could be produced by a criticality event that would thermally drive water out of the package at various percolation rates; and (3) provide an estimate of how the size of a package breach affects the water infiltration and therefore the possibility timing of criticality.

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Nodes packaging option for Space Station application

Space Station nodes packaging analyses are presented relative to moving environmental control and life support system (ECLSS) equipment from the habitability (HAB) module to node 4, in order to provide more living space and privacy for the crew, remove inherently noisy equipment from the crew quarter, retain crew waste collection and processing equipment in one location, and keep objectionable odor away from the living quarters. In addition, options for moving external electronic equipment from the Space Station truss to pressurized node 3 were evaluated in order to reduce the crew extravehicular-activity time required to install and maintain the equipment. Node size considered in this analysis is 3.66 m in diameter and 5.38 m long. The analysis shows that significant external electronic equipment could be relocated from the Space Station truss structure to node 3, and nonlife critical ECLSS HAB module equipment could be moved to node 4.

So, Kenneth T.↗

Leveraging Available Data to Support Extension of Transportation Packages Service Life

Data obtained from testing shipping package materials have been leveraged to support extending the service life of select shipping packages while in nuclear materials transportation. Increasingly, nuclear material inventories are being transferred to an interim storage location where they will reside for extended periods of time. Use of a shipping package to store nuclear materials in an interim storage location has become more attractive for a variety of reasons. Shipping packages are robust and have a qualified pedigree for their performance in normal operation and accident conditions within the approved shipment period and storing nuclear material within a shipping package results in reduced operations for the storage facility. However, the shipping package materials of construction must maintain a level of integrity as specified by the safety basis of the storage facility through the duration of the storage period, which is typically well beyond the one year transportation window. Test programs have been established to obtain aging data on materials of construction that are the most sensitive/susceptible to aging in certain shipping package designs. The collective data are being used to support extending the service life of shipping packages in both transportation and storage.

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Efficiencies and Optimization of Weak Base Anion Ion-Exchange Resin for Groundwater Hexavalent Chromium Removal at Hanford

The U.S. Department of Energy’s (DOE’s) contractor, CH2M HILL Plateau Remediation Company, has successfully converted a series of groundwater treatment facilities to use a new treatment resin that is delivering more than $3 million in annual cost savings and efficiency in treating groundwater contamination at the DOE Hanford Site in southeastern Washington State. During the production era, the nuclear reactors at the Hanford Site required a continuous supply of high-quality cooling water during operations. Cooling water consumption ranged from about 151,417 to 378,541 L/min (40,000 to 100,000 gal/min) per reactor, depending on specific operating conditions. Water from the Columbia River was filtered and treated chemically prior to use as cooling water, including the addition of sodium dichromate as a corrosion inhibitor. Hexavalent chromium was the primary component of the sodium dichromate and was introduced into the groundwater at the Hanford Site as a result of planned and unplanned discharges from the reactors starting in 1944. Groundwater contamination by hexavalent chromium and other contaminants related to nuclear reactor operations resulted in the need for groundwater remedial actions within the Hanford Site reactor areas. Beginning in 1995, groundwater treatment methods were evaluated, leading to the use of pump-and-treat facilities with ion exchange using Dowex™ 21K, a regenerable, strong-base anion exchange resin. This required regeneration of the resin, which was performed offsite. In 2008, DOE recognized that regulatory agreements would require significant expansion for the groundwater chromium treatment capacity. As a result, CH2M HILL performed testing at the Hanford Site in 2009 and 2010 to demonstrate resin performance in the specific groundwater chemistry at different waste sites. The testing demonstrated that a weak-base anion, single-use resin, specifically ResinTech SIR-700 ®, was effective at removing chromium, had a significantly higher capacity, could be disposed of efficiently onsite, and would eliminate the complexities and programmatic risks from sampling, packaging, transportation, and return of resin for regeneration.

54 ENVIRONMENTAL SCIENCES↗

MP197HB; First licensed transportation package with canistered high burnup fuel after dry storage - 15580

AREVA TN has recently received the NRC 10 CFR Part 71 transportation license CoC 9302 Revision 7 for the MP197HB transportation package that allows transportation of canistered high burnup fuel assemblies. The Nuclear Horizontal Modular Storage (NUHOMS{sup R}) system canisters containing high burnup PWR and BWR fuel assemblies are now licensed for transportation as a payload in MP197HB after these canisters were in dry storage at the utility ISFSIs. Enhancements have been made to the MP197HB transportation package to improve thermal performance and also provide superior shielding capabilities. Design features of existing AREVA TN transportation packages such as radial fins, thermal shield, neutron resin and optimized gamma and neutron shielding were used to accommodate transportation of high burnup fuel assemblies with heat loads up to 32 kW. The transportation package is designed for exclusive use by rail, truck or marine transport. State-of-the-art analyses methodologies were used to demonstrate compliance with regulatory requirements of 10 CFR 71.55 for the high burnup fuel after dry storage at any ISFSI. Some of the issues addressed in the license application related to the transportation of high burnup fuel included potential degradation of fuel cladding, uncertainty associated with high burnup fuel cladding material properties, and larger radiation and thermal sources due to higher burnups. Bounding conservative assumptions were made where limited test data was available. AREVA TN's licensing approach did not rely on the cladding integrity of high burnup fuel cladding. Instead, the criticality, thermal and shielding analyses were carried out assuming high burnup fuel reconfiguration. With the approval of this transportation package for PWR and BWR canistered high burnup fuel from dry storage at ISFSIs, the utilities can now transport these canisters which are in dry storage at their sites. This is also a major step forward for the canistered fuel that is currently in dry storage at the shutdown plants because they can use this AREVA TN MP197HB transportation package to ship that fuel to any interim storage facility (ISF) or to a repository when they become available. (authors)

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R and DE Robotic Sensor Intern Team: Remote 3D Mapping Sensor Suite

The primary goal of the sensor suite is to provide an internal 3D map of the HCAEX tunnel. When looking for a sensor that would provide XYZ point cloud data for mapping the decision was between a stereo camera, a 3D Lidar, or a 2D Lidar that would need to be spun around. The issues with the stereo camera were that it does not have a large field-of-view and that its depth information was not nearly as accurate as even the baseline Lidars. The selection between Lidars is detailed. The Hokuyo UTM-30LX-EW shown provides a 270 deg. FOV with high accuracy and resolution while the largest FOV on a 3D Lidar is 40 deg.. This was the reason a 2D Lidar was chosen over the 3D Lidar. Even though the 2D Lidar requires additional components and software to generate a 3D map, its FOV advantage is too great given that the HCAEX tunnel is a relatively confined space. The rotating planar scan from the 2D Lidar can be stitched into a 3D scan as shown in the graphic below. Along with XYZ data, the Lidar provides return intensity data in color which can be used to distinguish between different materials within the tunnel. To provide visual feedback a camera system needed to be identified. The common solution for this is a high resolution camera with a small FOV mounted on a pan-tilt system. But, given the critical nature of the visual feedback we looked into solutions without the need for moving parts. Two panospheric cameras give a FOV large enough to get a full 360 deg. view.. Camera color data (RGB) can also be combined with the 3D mapping data from the Lidar and overlaid on the map to give the user the ability to see inside the entire length of the tunnel as if they were inside. Situational awareness is required for the crawler in the tunnel in order for it to be able to register the 3D scans to create a final map. For this we procured a Microstrain 3DM-GX5-25 Inertial Measurement Unit (IMU). To perform the Lidar rotation we have two servos driving the pan and tilt. The servo controls and most data processing will be performed via an onboard computer. The sensor components are to be packaged together using 3D printed mounts into a standalone unit.

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Longevity of geo-synthetic cover components - 15450

Geo-synthetics form many of the key components in modern landfill cell and final cover designs. Models for the long-term performance assessment of systems incorporating these components have traditionally relied on a small number of standard assumptions originally developed to address the needs of the Resource Conservation and Recovery Act (RCRA) Subtitle D solid waste industry. These past estimates demonstrated that, even under a series of highly conservative factors, the longevity of these components far exceeds the requirements for typical solid waste facilities. Traditionally, the interpretation of the geo-synthetic degradation research has been deterministic as opposed to probabilistic, often due to the relative scarcity of degradation data for particular formulations and environmental conditions. Long-term studies of the mechanical and chemical properties of geo-synthetics subjected to accelerated aging processes have been conducted in an attempt to better quantify their longevity to address applications where performance may be required for 1,000 years or more. Of the many cover system geo-synthetic components, geo-membranes appear to have the most significant impact on the long-term performance of cover systems. Based on available long-term aging studies, a series of property degradation curves and their corresponding model parameters are available for geo-membranes. A strong area of need in the performance assessment of shallow disposal facilities is how to connect these research findings with inputs to probabilistic models. Significantly, factors affecting geo-membrane performance, such as geo-membrane thickness, anti-oxidant package, molecular weight, oxygen exposure, and drainage, can be adjusted and re-engineered using a proposed approach to meet specific performance targets. (authors)

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Activation Calculation for the Dismantling and Decommissioning of a Light Water Reactor Using MCNP{sup TM} with ADVANTG and ORIGEN-S

The boiling water reactor (BWR) Philippsburg 1 (KKP1) and the pressurized water reactor (PWR) Neckarwestheim 1 (GKN I), with a thermal power of 2575 MW{sub th} resp. 2497 MW{sub th}, were shut down and entered directly the post-operational phase which is covered by the operating license held by the nuclear power plant owner. The decommissioning of a nuclear power plant (NPP) which is licensed under chap. 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning requires a decommissioning license. This will include essential changes in requirements for the systems and components and will mainly lead to the immediate dismantling of the nuclear power plant which is now the only permitted decommissioning strategy in Germany. In this context, neutron activation calculations for the structural components of KKP1 and GKN I have to be carried out to predict occurring activities and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP6) combined with automated variance reduction tools (ADAVANTG) are used and are coupled with the activation and decay module ORIGEN-S. This paper presents the calculation procedure and the obtained results with a validation by measured activations and gamma dose rates.

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WTP Waste Feed Qualification: Hydrogen Generation Rate Measurement Apparatus Testing Report

The generation rate of hydrogen gas in the Hanford tank waste will be measured during the qualification of the staged tank waste for processing in the Hanford Tank Waste Treatment and Immobilization Plant. Based on a review of past practices in measurement of the hydrogen generation, an apparatus to perform this measurement has been designed and tested for use during waste feed qualification. The hydrogen generation rate measurement apparatus (HGRMA) described in this document utilized a 100 milliliter sample in a continuously-purged, continuously-stirred vessel, with measurement of hydrogen concentration in the vent gas. The vessel and lid had a combined 220 milliliters of headspace. The vent gas system included a small condenser to prevent excessive evaporative losses from the sample during the test, as well as a demister and filter to prevent particle migration from the sample to the gas chromatography system. The gas chromatograph was an on line automated instrument with a large-volume sample-injection system to allow measurement of very low hydrogen concentrations. This instrument automatically sampled the vent gas from the hydrogen generation rate measurement apparatus every five minutes and performed data regression in real time. The fabrication of the hydrogen generation rate measurement apparatus was in accordance with twenty three (23) design requirements documented in the conceptual design package, as well as seven (7) required developmental activities documented in the task plan associated with this work scope. The HGRMA was initially tested for proof of concept with physical simulants, and a remote demonstration of the system was performed in the Savannah River National Laboratory Shielded Cells Mockup Facility. Final verification testing was performed using non-radioactive simulants of the Hanford tank waste. Three different simulants were tested to bound the expected rheological properties expected during waste feed qualification testing. These simulants were tested at different temperatures using purge gas spiked with varying amounts of hydrogen to provide verification that the system could accurately measure the hydrogen in the vent gas at steady state.

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Effects of Copolymer Structure on Enzyme-Catalyzed Polyester Recycling

Polyesters are an omnipresent material used for a variety of applications (e.g., bottles, packaging, textile, windshield), roughly comprising 8% of plastics produced worldwide. Enzymatic recycling is an emerging solution to deal with the increasingly diverse polyesters that are not suitable for mechanical recycling. However, enzyme activity and efficiency are still the limiting factors impeding enzymatic recycling for different plastic waste forms. The effects of thermal and structural properties (e.g., glass transition temperature, crystallinity, specific surface area), which are determined by chemical composition and preprocessing, directly influence enzyme recycling efficiency. This work investigates two extrusion methods (single screw and twin-screw extrusion) to pretreat a range of copolyesters (RPET, PETG, Ecozen, Tritan and PBT) and modify their properties (i.e., glass transition temperature (T g ), crystallinity (%), and molecular weight (M n )). A PET-specific enzyme, leaf-branch compost cutinase (LCC ICCG ), produced from a fed-batch fermentation of Escherichia coli BL21(DE3), was used for the enzymatic depolymerization of different polyesters. Several copolyesters showed improved depolymerization after pretreatment, as measured by rate and amount of monomers produced. Furthermore, those that did not depolymerize were found to have exceptionally high glass transition temperature or percent crystallinity, highlighting the importance of these physical parameters on conversion efficiency.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Spent Nuclear Fuel Disposition

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNF is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Standards and Guidelines for the Design of Trunnions of RAM Transport Packages - 17365

Trunnion systems of packages for the transport of radioactive materials have to guarantee the safe handling of package during crane operations (lifting, tilting) and to secure package tie down to the transport vehicle, if the trunnions are used as attachment points during transport. The design of trunnions is based on IAEA Regulations SSR-6, the supporting Advisory Material SSG-26 and further appropriate technical standards and/or guidelines. For package approval procedures in Germany the guideline BAM-GGR 012 has to be applied. If a package is handled inside a german nuclear power plant, the standard KTA 3905 has to be fulfilled additionally. In this paper the requirements of KTA 3905 [4] concerning the trunnion systems as load attaching points (LAP) are discussed in connection with the recommendations in the guideline BAM-GGR 012 [3]. This guideline is prepared at BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems of Type B(U) transport packages. The quality assurance questions concerning trunnion systems are discussed as well. (authors)

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VENTSAR V3.0: A Python GUI for Estimating Contaminant Concentrations on or Near Buildings Due to Building Effects and Plume Rise and For Calculating Inhalation and Plume Shine Doses

VENTSAR: Originated as VENTAX (Smith and Weber 1983) on the IBM Mainframe at the Savannah River Site (SRS) as a Fortran program. Updated to VENTSAR XL V1.0 (Simpkins 1997) as a spreadsheet version using macros created in Microsoft Excel. Later updated to VENTSAR XL V2.0 (Dixon 2018) as a spreadsheet executing a modified version of the original VENTAX Fortran code. Estimates contaminant concentrations on or near a building from a release at a nearby location. Calculates concentrations for a given meteorological exceedance probability or for a given stability and wind speed combination. Can model a single building with or without a penthouse on top or a ground location from either a stack or ground release. Plume rise can be considered. Contaminant releases can be chemical or radioactive with downwind concentrations determined at user-specified distances. Wind passing over and around buildings creates a complicated dispersion pattern. Air-intake vents may be located on building roofs or near the ground downwind of a release source. An estimation of pollutant concentrations on or near a structure is important in determining expected pollutant levels. Meteorological data are selected based on a specific area of the site. Fortran-based VENTAX able to make fast, complex mathematical calculations. Not user-friendly VENTSAR XL V1.0 no longer supported due to obsolete Excel macros. VENTSAR XL V2.0 an interim solution using an Excel spreadsheet to interface with the modified VENTAX Fortran code. Requires user to have Microsoft Excel. Not an intuitive interface for someone unfamiliar with VENTSAR. Does not perform dose calculations. VENTSAR V3.0 goal to combine the strengths of previous versions of VENTSAR into a single, powerful yet user-friendly program with a Graphical User Interface (GUI). Not dependent upon a specific program or plug-ins. Self-contained package to be used on any computer running Microsoft Windows. User-friendly, intuitive GUI created in Python for parameter input. Executes same modified VENTAX Fortran code as VENTSAR XL V2.0. Outputs formatted text file of completed calculations for easy review and dissemination. Performs inhalation and plume shine dose calculations for up to 11 user-selected nuclides from a nuclide dose factor library of almost 500 nuclides. 12 test cases were created for verification of V1.0 and V2.0. Same test cases run in V3.0 to verify correct performance. V3.0 produced similar results to V1.0. Confirmed GUI did not alter calculations in any way. Comparisons of the test case results confirm the V3.0 GUI does not influence the VENTSAR calculations. Simply passes same input parameters to VENTAX Fortran code for execution. Provides end user an easy-to-use, intuitive tool to quickly make building effect and plume rise calculations, as well as, inhalation and plume shine dose calculations. No experience with or working knowledge of Fortran, Excel, command line, or macros required. Self-contained package allows VENTSAR be deployed to any Windows computer without the need for specific programs or plug-ins.

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Coupling the Modeled Structural Transmissibility of a Used Nuclear Fuel Conveyance to Over the Road Data - 16215

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Current information is insufficient to determine the ability of UNF, including high-burnup fuel, to withstand shock and vibration loads that could occur when UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility after extended storage. In order to make this determination, the magnitude of the transportation loads transmitted to the UNF must be quantified. Previous preliminary modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads and the importance of modeling all aspects of the transport system (i.e. rail car, transport cradle, cask, canister, and fuel). The work presented herein proposes a methodology for determining the structural transmissibility of a hypothetical transport system, and conditions existing over the road test (OTR) data to predict the performance of the hypothetical system. This method utilizes the modeled frequency response of the hypothetical system to scale the existing OTR acceleration data. This work will be relevant in creating models of a UNF conveyance during transport. As such, preliminary work in coupling the rail vehicle dynamics code NUCARS{sup C}, a general purpose multi-body rail vehicle dynamics simulation package developed by Transportation Technology Center Inc. (TTCI), to existing models of UNF during transport are also discussed herein. (authors)

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Mechanical and chemical recycling of solid plastic waste

Highlights: • This paper reviews the current pathways for recycling of solid plastic waste, via both mechanical and chemical recycling. • The predominant industrial technologies, design strategies and recycling examples of specific waste streams are reviewed. • The main challenges and some potential remedies are discussed. - Abstract: This review presents a comprehensive description of the current pathways for recycling of polymers, via both mechanical and chemical recycling. The principles of these recycling pathways are framed against current-day industrial reality, by discussing predominant industrial technologies, design strategies and recycling examples of specific waste streams. Starting with an overview on types of solid plastic waste (SPW) and their origins, the manuscript continues with a discussion on the different valorisation options for SPW. The section on mechanical recycling contains an overview of current sorting technologies, specific challenges for mechanical recycling such as thermo-mechanical or lifetime degradation and the immiscibility of polymer blends. It also includes some industrial examples such as polyethylene terephthalate (PET) recycling, and SPW from post-consumer packaging, end-of-life vehicles or electr(on)ic devices. A separate section is dedicated to the relationship between design and recycling, emphasizing the role of concepts such as Design from Recycling. The section on chemical recycling collects a state-of-the-art on techniques such as chemolysis, pyrolysis, fluid catalytic cracking, hydrogen techniques and gasification. Additionally, this review discusses the main challenges (and some potential remedies) to these recycling strategies and ground them in the relevant polymer science, thus providing an academic angle as well as an applied one.

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Preliminary Concept of Operations for the Spent Fuel Management System+ - 17134

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the US Department of Energy's Office of Nuclear Energy is tasked with identifying, planning, and conducting activities for developing interim storage and transportation capabilities in support of an integrated waste management system (IWMS). The IWMS is expected to transport Spent Nuclear Fuel (SNF) from a reactor and deliver it to a repository in a suitable container, with the possibility of interim storage. The system will be composed of multiple subsystems including possible interim storage facilities (ISFs), one or more repositories, facilities to package and/or repackage SNF, and transportation. The project team is analyzing options for an IWMS to identify viable options. To support analysis, the project team has developed a concept of operations framework for SNF related components of the potential integrated system and the interdependencies between those components. A goal of this work is to aid system analysts in developing consistent models across the project. The concept of operations framework will be updated periodically as new developments emerge. At a high level, SNF is expected to be transported from reactor sites to a repository. Initially, SNF is unloaded from reactors and placed in spent fuel pools (SFPs) for wet storage at utility sites. To maintain space in SFPs, SNF is placed in containers at reactor sites for storage and/or transportation only after the SNF has cooled enough to satisfy loading limits. Once transportation requirements are met, the SNF may be transported to an ISF, where it is stored until a repository is developed, or it is transported directly to a repository if one is available.While the high level operation of the system is straightforward, analysts must evaluate numerous alternatives, such as the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if needed), repackaging technology, types of containers used, repository design, component sizing, and timing of events. These alternatives arise due to technological, economic, or policy considerations. This paper provides an overview of the potential alternatives identified in the concept of operations framework at a conceptual level. (authors)

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L-SCIE and SUPCRTNE model and database development

This progress report (Level 3 Milestone Number M3SF-25LL010301052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Host Rock Properties & Processes Work Package SF-25LL01030105. The focus of this milestone is to create an initial SUPCRTNE database and extensively test the SUPCRTNE code. We prepared a draft manuscript describing our workflow and approach for developing a next-generation thermodynamic database for the Spent Fuel and High Level Waste Disposition campaign. The goal is to ease any future formal software qualification effort as was done on the Yucca Mountain Project for codes including SUPCRT92 and EQ3/6. We are attempting to extend the SUPCRTNE database to include more species, mainly from the NEA volumes. We can also take advantage of thermodynamic database activities that are ongoing at Thermochimie and as part of the EURADII program. However, our effort is limited by the funds available and the changes to program scope that was initiated in mid-FY25.

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Crystalline Disposal R&D at LBNL: FY20 Progress Report

Within the Spent Fuel & Waste Science and Technology (SFWST) Program, research work continues further with the goal to better understand long-term performance of disposal systems in three main geologic rock types: clay/shale, salt, and crystalline rock. This report documents LBNL’s research activities related to investigations of crystalline host rock according to the scope of two work packages: SF-20LB01030207 “Crystalline International Collaborations – LBNL,” and SF-20LB01030203 “Crystalline Disposal R&D – LBNL.” These research activities correspond are related to key Features, Events, and Processes (FEPs), ranked in importance from medium to high, as listed in Table 7 of the Used Fuel Disposition Campaign Disposal Research and Development Roadmap (FCR&D-USED-2011-000065 REV0) (Nutt, 2011). Specifically, these research activities address FEP 2.2.01, Excavation Disturbed Zone (EZD). The results of these research activities provide important insights into understanding and predicting flow and transport processes that could occur in low-permeability crystalline rocks, in which fractures might serve as main conduits for fluid flow and radionuclide transport. The evolution of the EDZ during the excavation of the tunnel as well as the evolution of microcrack growth within the EDZ after the emplacement of backfill are critical for predicting the long- term behavior of the EDZ. A number of factors including stress, temperature, water activity, capillary pressure, chemistry, and mineralogy can affect the rock evolution, which require advanced experimental tools to study it.

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