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Uranyl oxalate species in high ionic strength environments: stability constants for aqueous and solid uranyl oxalate complexes

Uranyl ion, UO 2 2+ , and its aqueous complexes with organic and inorganic ligands can be the dominant species for uranium transport on the Earth surface or in a nuclear waste disposal system if an oxidizing condition is present. As an important biodegradation product, oxalate, C 2 O 4 2– , is ubiquitous in natural environments and is known for its ability to complex with the uranyl ion. Oxalate can also form solid phases with uranyl ion in certain environments thus limiting uranium migration. Therefore, the determination of stability constants for aqueous and solid uranyl oxalate complexes is important not only to the understanding of uranium mobility in natural environments, but also to the performance assessment of nuclear waste disposal. In this work, we developed a thermodynamic model for the UO 2 2+ –Na + –H + –Cl – –ClO 4 – –C 2 O 4 2– –NO 3 – –H 2 O system to ionic strength up to ~11 mol•kg –1 . We constrained the stability constants for UO 2 C 2 O 4 (aq) and UO 2 (C 2 O 4 ) 2 2– at infinite dilution based on our evaluation of the literature data over a wide range of ionic strengths up to ~11 mol•kg –1 . We also obtained the solubility constants at infinite dilution for solid uranyl oxalates, UO 2 C 2 O 4 •3H 2 O, based on the solubility data over a wide range of ionic strengths. The developed model will enable for the accurate stability assessment of oxalate complexes affecting uranium mobility under a wide range of conditions including those in deep geological repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

HTO and selenate diffusion through compacted Na-, Na–Ca-, and Ca-montmorillonite

Radionuclide transport in smectite clay barrier systems used for nuclear waste disposal is controlled by diffusion, with adsorption significantly retarding transport rates. While a relatively minor component of spent nuclear fuel, 79 Se is a major driver of the safety case for spent fuel disposal due to its long half-life (3.3×10 5 yr) and its low adsorption to clay (K D < 10 L/kg), thus a thorough understanding of Se diffusion through clay is critical for understanding the long-term safety of spent fuel disposal systems. Through-diffusion experiments with tritiated water (HTO, conservative tracer) and Se(VI) were conducted with a well-characterized, purified montmorillonite source clay (SWy-2) under a constant ionic strength (0.1 M) and three different electrolyte compositions: Na + , Ca 2+ , and a Na + -Ca 2+ mixture at pH 6.5 in order to probe the effects of electrolyte composition and interlayer cation composition on clay microstructure, Se(VI) aqueous speciation, and ultimately diffusion. Further, the results were modeled using a reactive transport modeling approach to determine values of porosity (ε), D e (effective diffusion coefficient), and K D (distribution coefficient for adsorption). HTO diffusive flux was higher in Ca-montmorillonite (D e =1.68×10 -10 m 2 s -1 ) compared to Na-montmorillonite (De=7.83×10 -11 m 2 s -1 ). This increase in flux is likely due to a greater degree of clay layer stacking in the presence of Ca 2+ compared to Na + , which leads to larger inter-particle pores. Overall, the Se(VI) flux was much lower than the HTO flux due to anion exclusion, with Se(VI) flux following the order Ca (D e = 1.03×10 -11 m 2 s -1 ) > Na–Ca (D e = 2.12×10 -12 m 2 s -1 ) > Na (D e = 1.28×10 -12 m 2 s -1 ). These differences in Se(VI) flux are due to a combination of factors, including (1) larger accessible porosity in Ca-montmorillonite due to clay layer stacking and smaller electrostatic effects compared to Na-montmorillonite, (2) larger accessible porosity for neutral-charge CaSeO4 species which makes up 32% of aqueous Se(VI) in the pure Ca system, and (3) possibly higher Se(VI) adsorption for Ca-montmorillonite. Through a combination of experimental and modeling work, this study highlights the compounding effects that electrolyte and counterion compositions can have on radionuclide transport through clay. Diffusion models that neglect these effects are not transferable from laboratory experimental conditions to in situ repository conditions.

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Comparative modelling of the coupled thermal–hydraulic-mechanical (THM) processes in a heated bentonite pellet column with hydration

For the deep geological disposal of high-level radioactive waste in argillaceous rocks, the heat production of the waste is an important driver for thermal–hydraulic-mechanical (THM)-coupled processes. These THM processes influence the properties and conditions of the near field that in many repository designs contains bentonite as a clay buffer. One task in the DECOVALEX-2015 (DEvelopment of COupled models and their VALidation against Experiments) project was the modelling of a heated bentonite column (Villar et al. in Long-term THM tests reports: THM cells for the HE-E test: update of results until February 2014. Deliverable-no: D2.2-7.3. CIEMAT Technical Report IEMAT/DMA/2G210/03/2014, 2014) in preparation for the in situ heater experiment HE-E at the underground rock laboratory Mont Terri. DECOVALEX is an international cooperative project that focuses on the development and validation of mathematical models for simulating such coupled processes associated with disposal in deep geological repositories. Eight modelling teams developed their own THM-coupled models for the bentonite column experiment, using six different simulation codes. Each of the teams individually calibrated the THM parameters for the bentonite material. The eight resulting parameter sets agree well and allow a satisfactory reproduction of the TH measurements by all models. The modelling results for the evolution of temperature and relative humidity over time at three sensors in the bentonite column are in good agreement between the teams and with the measured data. Also, changes of the temperature due to modifications of the insulation and the adjustment of the heating power during the course of the experiment are well reproduced. The models were thus able to reproduce the main physical processes of the experiment, both for vapour-dominated diffusion during the heating phase and combined liquid and vapour transport during a subsequent heating and hydration phase. Lastly, based on the parameter sets, the teams predict a penetration of the water infiltration front in the 48-cm column filled with bentonite pellets to a depth between 25 and 35 cm over the 15,000 h (i.e. over 20 months) of the hydration phase of the experiment.

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Proceedings of the 6th US/German Workshop on Salt Repository Research, Design, and Operation

The 6th US/German Workshop on Salt Repository Research, Design, and Operation was held in Dresden. Germany on September 7-9, 2015. Over seventy participants helped advance the technical basis for salt disposal of radioactive waste. The number of collaborative efforts continues to grow and to produce useful documentation, as well as to define the state of the art for research areas. These Proceedings are divided into Chapters, and a list of authors is included in the Acknowledgement Section. Also in this document are the Technical Agenda, List of Participants, Biographical Information, Abstracts, and Presentations. Proceedings of all workshops and other pertinent information are posted on websites hosted by Sandia National Laboratories and the Nuclear Energy Agency Salt Club. The US/German workshops provide continuity for long-term research, summarize and publish status of mature areas, and develop appropriate research by consensus in a workshop environment. As before, major areas and findings are highlighted, which constitute topical Chapters in these Proceedings. In total, the scientific breadth is substantial and while not all subject matter is elaborated into chapter format, all presentations and abstracts are published in this document. In the following Proceedings, six selected topics are developed in detail.

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M4SF-19LL010301042: Code Development in Support of the SFWST Thermodynamic Database

This progress report (Level 4 Milestone Number M4SF-19LL010301042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Activity Number SF-19LL010301041. LLNL is leading efforts in the development of thermodynamic databases in support of the Spent Fuel and Waste and Science Technology (SFWST) program. Thermodynamic models provide the basis for understanding the stability of solid phases and speciation of aqueous species and modeling the evolution of repository conditions. The LLNL effort is being performed in coordination with other US database development efforts. The effort includes a review of available thermochemical databases and a path forward for database integration. International coordination with the NEA-TDB is supported through crystalline international work package

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

There Are Safe Limits for Radioactivity - Does WIPP Need to Filter its Exhaust Air? - 17165

Pilot Plant (WIPP) underground repository resulted in the release of americium and plutonium (mostly {sup 241}Am) into the environment and contaminated portions of the repository, primarily along the ventilation path from the location of the incident. The WIPP uses a ventilation system to ensure that underground air is circulated through the repository and to ensure that air quality conditions are safe for the workers. Since the air in the repository exits to the surface through an exhaust shaft, this shaft is the sole potential pathway for airborne radioactivity release from the repository. During normal operations before the 2014 incident, exhaust air was released to the environment unfiltered, with an automatic switch to filtration when airborne radioactivity was detected immediately down-stream from the active waste disposal operation. However, since the radiation release event, the WIPP ventilation system has been, and will likely remain in what is called 'filtration mode,' (i.e., exhaust air is routed through a HEPA filter system). Redirection of the ventilation system through the HEPA (High Efficiency Particulate Air) filter system was necessary at the peak of the radiation release event to protect aboveground workers at the site and the public in the surrounding areas; however, it has hampered recovery efforts and has exacerbated the inherent safety issues of working underground. As a component of the Carlsbad Environmental Monitoring and Research Center's (CEMRC) WIPP environmental monitoring program, airborne radioactivity samples are collected daily in the exhaust shaft ventilation air as it reaches the surface, and before entering the HEPA filters. Since June 2014, airborne radiation levels entering the HEPA filters have averaged about 1% of a derived air concentration (DAC) for {sup 241}Am. An airborne radioactivity level of one DAC is equivalent to a potential worker internal dose exposure of 50 mSv per work year, if a worker were continuously exposed for a full work year (2000 hours). During Project Reach and throughout decontamination campaigns conducted in Panel 7 for several months in 2015, these unfiltered levels rose to 3-5% of a DAC, with a one week spike reaching almost one DAC in June 2015, during decontamination of the most heavily contaminated exhaust drift in Panel 7. Since that period, the unfiltered levels have never exceeded 2% of a DAC. Going forward, the question is, should WIPP resume unfiltered discharge of its underground ventilation air? Evaluation of more than two years of underground ventilation sampling data indicates that residual radioactivity levels in the underground no longer warrant HEPA filtration in order to meet either worker or environmental protection criteria. In terms of radiological risk at or in the vicinity of the WIPP site, the increased risk from resuming unfiltered ventilation is exceedingly small and would not result in a hypothetical dose to a member of the public in excess of the 100 micro Sv per year limit for periodic confirmatory sampling as required for operating repositories by the Environmental Protection Agency (40 CFR 191). In addition, the requirement for respiratory protection for workers underground exposed to 1-2% of a DAC is questionable, given the added industrial safety risk of heavy equipment operation while wearing respirators. (authors)

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M4SF-17LL010301071: Thermodynamic Database Development

This progress report (Level 4 Milestone Number M4SF-17LL010301071) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Disposal R&D Work Package Number M4SF-17LL01030107. The DR Argillite Disposal R&D control account is focused on the evaluation of important processes in the analysis of disposal design concepts and related materials for nuclear fuel disposal in clay-bearing repository media. The objectives of this work package are to develop model tools for evaluating impacts of THMC process on long-term disposal of spent fuel in argillite rocks, and to establish the scientific basis for high thermal limits. This work is contributing to the GDSA model activities to identify gaps, develop process models, provide parameter feeds and support requirements providing the capability for a robust repository performance assessment model by 2020.

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Status of Multiphysics Modeling of Critical DPCs in a Geological Repository

A summary of previous work focused on the first two scenarios is presented as an established workflow. These canisters could reach an equilibrium at a low quasi–steady-state power level that could be sustained for thousands of years. The majority of this report describes an effort to expand the initial methodology to include the scenario in which a DPC has enough excess reactivity to boil the water inside the canister. This scenario involves a variety of tools and modeling approaches to capture the two-phase behavior, including RELAP5-3D, STAR-CCM+, and TRACE. Efforts using the high-fidelity computational fluid dynamics results to verify the accuracy of RELAP5-3D and TRACE for boiling DPCs are also presented. A new modeling strategy using TRACE appears to be the most promising approach for expanding the analysis of critical DPCs beyond the subcooled scenarios. Finally, this report describes ongoing work investigating the fourth scenario, in which a critical DPC is sealed in saturated bentonite. The primary goal in this research is to understand the power output and associated pressure increase to inform geomechanics simulations that could predict the impact on a geological repository.

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Initial Evaluation of Microreactor Disposition Options

The United States Department of Energy is supporting the U.S advanced reactor industry through funding, legislation and regulatory development to actively pursue several microreactor design concepts. The Idaho National Laboratory (INL) is strategically positioned to support demonstration of microreactor technology in the next three to five years. This report provides and initial evaluation of the disposition options for microreactor spent nuclear fuel (SNF) generated as part of the microreactor technology demonstration program. Currently available information constitutes the basis for the options identified and discussions thereof. In the absence of detailed microreactor design information, assumptions were made to facilitate the identification of disposition options Particulars pertaining to any component of an identified disposition pathway are naturally highly microreactor design specific and in general such details are not provided. The diverse nature of potential microreactor SNF is reflected in the diverse nature of DOE owned SNF stored at INL. Therefore, it is anticipated that DOE currently stores and manages fuels that can serve as analogs for most microreactor fuel concepts. As such disposition options for microreactor SNF are expected to be much the same as that for these existing fuels. Two generic microreactor concepts have been selected for the purposes of this options assessment. The selected reactor concepts are a tristructural isotropic (TRISO) fueled high temperature gas reactor concept and a sodium/potassium bonded heat pipe reactor with uranium oxide fuel. Both concepts are assumed to be using high-assay low enriched uranium (HALEU) as the initial fuel composition. Interim storage, treatment and neutralization, material recovery, packaging and extended dry storage options are identified these reactors. The disposition options include existing INL facilities and capabilities and new facilities and capabilities developed as part of the microreactor program or as part of DOE's overall strategy for the eventual transfer of all SNF at INL to a permanent repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Using Radioiodine Speciation to Address Environmental Remediation and Waste Stream Sequestration Problems at the Fukushima Daiichi Nuclear Power Plant and a DOE Site

Iodine-129 ( 129 I), with a half-life of half-life of 16 million years, is commonly considered the single greatest risk driver in high-level and low-level nuclear repositories. This risk stems from several basic properties of 129 I, and under many geochemical conditions, it can move as an anion at nearly the rate of water through the subsurface environment. 129 I is also extremely radiologically toxic because over 90% of body burden accumulates in the thyroid, which weighs only about 14g in an adult. There is also a large worldwide inventory of radioiodine as a result of its high fission yield and this inventory is rapidly increasing as a result of nuclear energy production. Radioiodine is produced at a rate of 40 GBq (1 Ci) per gigawatt of electricity produced by nuclear power. To illustrate how the properties of 129 I magnify its risk, 129 I accounts for only 0.00002% of the radiation released from the Savannah River Site in Aiken, South Carolina, but contributes 13% of the population dose, a six orders of magnitude magnification of risk with respect to its radioactivity. The currently favored solid phase for LLW immobilization is cement, while HLW immobilization is the incorporation of waste into glass (vitrification). However, so far, the incorporation into cement and subsequent leaching of only iodide has been seriously investigated. The major problem with this approach is that it ignores the complex speciation of iodine, i.e., it ignores iodate and organo-iodine which have different chemistries. Most of the past research was devoted to the mechanism of iodide uptake in cement hydrate phases that is sorption and/or incorporation. Very few data exist on iodate and organo-iodine incorporation in cement, even though large quantities of liquid waste containing also radioiodine have already been solidified in cement Iodine-129 from low-level waste is commonly disposed of in cementitious materials. Grout, a dense cementitious fluid, mixed with a reducing slag, is often used to immobilize radionuclides. However, the reducing environment might not be conducive to immobilize iodine. For example, the silver based immobilization technologies (e.g., AgCl, Ag-impregnated granular activated carbon, Ag-mordenite) remove iodine from the aqueous phase by promoting the formation of Ag-iodide precipitates. The solubility of AgI is eight orders of magnitude lower than it is for AgIO 3 . Similarly, coprecipitation of iodine into calcium carbonate phases occurs only with IO 3 - and not with I - and org-I. If one would want to immobilize iodine more effectively, different engineering approaches would need to be used to promote binding of I - , IO 3 - , or organo-I. Using laboratory experiments with grout, slag, and silver-based adsorbents, and GC-MS and I K-edge XANES and EXAFS and C K-edge XANES spectroscopy for identifying iodine speciation, the major problems with these methods have been identified as focused too much on just one of the iodine species for immobilization, while others, especially organo-I, remained mobile. Finally, we established that most of the adsorbents that are used contain sufficient amounts of organic matter to create organo-I . It is anticipated that increased attention directed at understanding and quantifying the speciation of radioiodine, as opposed to simply total radioiodine, will lead to improved remediation results to be used for long-term radioiodine disposal in cementitious waste forms.

129-Iodine↗

Prospects and Challenges in the Management of Non-Retrievable Radioactive Waste in the Russian Federation - 17212

In 2011, the Federal Law 'On the Management of Radioactive Waste' was enacted in the Russian Federation setting a framework for the Unified State System for Radioactive Waste Management and establishing a mandatory requirement for safe radioactive waste (RW) disposal. Particular category of 'non-retrievable RW' was defined given the amount of RW generated by that time, the issues associated with relevant sites, as well as the experience gained during the implementation of nuclear legacy programs. It was specified that in situ disposal is the preferred management option for this category of waste rather than RW retrieval and re-disposal due to relevant radiation and financial aspects. In 2011 - 2015, a set of high-priority tasks on taking the inventory of accumulated RW was accomplished. This work also provided for demonstrating the feasibility and making key decisions on the final stages on their lifecycle and relevant end-state strategies. Facilities holding non-retrievable RW were divided into two categories: - subject to transition into near-surface storage facilities by constructing additional safety barriers; - subject to controlled long-term storage followed by remediation. Efforts on a number of surface liquid RW storage water reservoirs falling under the first category launched back in the 20. century were completed. The new federal program provides for a set of follow-up efforts to be completed at 12 facilities. As it comes to the second category, the main task was to develop a set of safety management options, as well as some computational methods to demonstrate the feasibility of the selected end-state strategies. The most considerable efforts were made at the Techa Cascade of water reservoirs being the largest liquid radioactive waste (LRW) storage facility in the world that holds over 360 million m{sup 3} of waste with uncertain legal status, the lack of relevant computational models to predict its state. The paper summarizes the data on defining the end states of non-retrievable RW, completed and scheduled efforts on implementing relevant long-term safety strategies. The paper focuses on subsurface repositories and near-surface LRW storage reservoirs considered as common types of facilities holding non-retrievable RW. (authors)

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Status of Progress Made Toward Preliminary Design Concepts for the Inventory in Select Media for DOE-Managed HLW/SNF

As the title suggests, this report provides a summary of the status and progress for the Preliminary Design Concepts Work Package. Described herein are design concepts and thermal analysis for crystalline and salt host media. The report concludes that thermal management of defense waste, including the relatively small subset of high thermal output waste packages, is readily achievable. Another important conclusion pertains to engineering feasibility, and design concepts presented herein are based upon established and existing elements and/or designs. The multipack configuration options for the crystalline host media pose the greatest engineering challenges, as these designs involve large, heavy waste packages that pose specific challenges with respect to handling and emplacement. Defense-related Spent Nuclear Fuel (DSNF) presents issues for post-closure criticality control, and a key recommendation made herein relates to the need for special packaging design that includes neutron-absorbing material for the DSNF. Lastly, this report finds that the preliminary design options discussed are tenable for operational and post-closure safety, owing to the fact that these concepts have been derived from other published and well-studied repository designs.

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Integration of the Back End of the Nuclear Fuel Cycle

Management of spent nuclear fuel and high-level radioactive waste consists of three main phases – storage, transportation, and disposal – commonly referred to as the back end of the nuclear fuel cycle. Current practice for commercial spent nuclear fuel management in the United States (US) includes temporary storage of spent fuel in both pools and dry storage systems at operating or shutdown nuclear power plants. Storage pools are filling to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler spent fuel from pools into dry storage. Unless a repository becomes available that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 136,000 metric tons of spent fuel in dry storage systems by mid-century, when the last plants in the current reactor fleet are decommissioned. Current designs for dry storage systems rely on large multi-assembly canisters, the most common of which are so-called “dual-purpose canisters” (DPCs). DPCs are certified for both storage and transportation, but are not designed or licensed for permanent disposal. The large capacity (greater number of spent fuel assemblies) of these canisters can lead to higher canister temperatures, which can delay transportation and/or complicate disposal. This current management practice, in which the utilities continue loading an ever-increasing inventory of larger DPCs, does not emphasize integration among storage, transportation, and disposal. This lack of integration does not cause safety issues, but it does lead to a suboptimal system that increases costs, complicates storage and transportation operations, and limits options for permanent disposal. This paper describes strategies for improving integration of management practices in the US across the entire back end of the nuclear fuel cycle. The complex interactions between storage, transportation, and disposal make a single optimal solution unlikely. However, efforts to integrate various phases of nuclear waste management can have the greatest impact if they begin promptly and continue to evolve throughout the remaining life of the current fuel cycle. A key factor that influences the path forward for integration of nuclear waste management practices is the identification of the timing and location for a repository. The most cost-effective path forward would be to open a repository by mid-century with the capability to directly dispose of DPCs without repackaging the spent fuel into disposalready canisters. Options that involve repackaging of spent fuel from DPCs into disposalready canisters or that delay the repository opening significantly beyond mid-century could add 10s of billions of dollars to the total system life cycle cost.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Summary of FY15 Results of Benchmark Modeling Activities

Sandia is participating in the third phase of and is a contributing partner to a U.S.- German “Joint Project” entitled “Comparison of current constitutive models and simulation procedures on the basis of model calculations of the thermo-mechanical behavior and healing of rock salt.” The first goal of the project is to check the ability of numerical modeling tools to correctly describe the relevant deformation phenomena in rock salt under various influences. Achieving this goal will lead to increased confidence in the results of numerical simulations related to the secure storage of radioactive wastes in rock salt, thereby enhancing the acceptance of the results. These results may ultimately be used to make various assertions regarding both the stability analysis of an underground repository in salt, during the operating phase, and the long-term integrity of the geological barrier against the release of harmful substances into the biosphere, in the post-operating phase. Among the numerical modeling tools required to address this are constitutive models that are used in computer simulations for the description of the thermal, mechanical, and hydraulic behavior of the host rock under various influences and for the long-term prediction of this behavior into the future. A second goal of the project is to investigate and demonstrate the possibilities for further potential development and improvement of these constitutive models. This report summarizes the efforts undertaken during FY15 in support of this international U.S.-German benchmark initiative.

58 GEOSCIENCES↗

Regional Geologic Evaluations for Disposal of HLW and SNF: Alluvial Basins of the Basin and Range Province

The DOE Spent Fuel and Waste Technology (SFWT) R&D Campaign supports research on crystalline rock, shale (argillite) and salt as potential host rocks for disposal of HLW and SNF in a mined geologic repository. Modeling involving these host rocks for the generic GDSA reference cases has assumed saturated conditions at repository depths. In order to test two-phase flow capabilities of the GDSA flow and transport models, the SFWT Campaign is developing an unsaturated zone reference case. Alluvial basin-fill deposits within the Basin and Range province have unsaturated zones that can be as thick as 600 meters, although the thickness is typically much less. Basin-fill alluvium has been considered as a potential medium for HLW disposal in the US since the 1950s (NAS,1957; Smyth et al., 1979; Winograd, 1981). Attributes considered favorable for HLW disposal in alluvium in the Basin and Range province include an arid environment, low water flux , high sorption capacity (from clay s and zeolites), and ease of construction and stability of repository openings (Hardin et al., 2012). Hardin et al. (2012) also address the thermal issues inherent in the design of a HLW repository in alluvium with low thermal conductivity, which are not discussed in this report.

58 GEOSCIENCES↗

(Project 18-15502) Reducing Uncertainty in Radionuclide Transport Prediction Using Multiple Environmental Tracers (Final Report)

In order to successfully site and design a nuclear waste disposal facility, DOE scientists are required to show safe containment of the radioactive waste for up to 1 million years. A significant hurdle to accurately and convincingly demonstrating disposal safety is predicting the fate of radioactive elements once they enter the groundwater system surrounding the repository. These predictions are often made with computer models, which contain accurate physics of groundwater movement and chemical reactions that occur during groundwater flow. As computer power increases, the physics and chemistry of these computer simulators can become more and more realistic and the physically based error decreases. However, the ability of these computer models to provide accurate predictions in a specific place, over long time periods, requires the scientists and engineers to know the subsurface properties of the Earth that control groundwater movement. In particular, groundwater scientists and engineers need to know the groundwater fluid velocity, which can change over time and strongly vary with location within the groundwater system. Because the Earth’s subsurface cannot be directly seen, and can only be sampled at drilling locations, the properties of the groundwater system are never known completely, and computer models of groundwater transport will always have some amount of uncertainty. This project’s principal goal was to use chemicals and isotope “tracers”, which have been introduced to the groundwater system by natural processes over long time periods, to help inform computer models of the groundwater velocity and subsurface properties. The goal was to calculate how much better predictions of groundwater transport were when these “tracers” were used to inform the computer models.

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The Next Generation System Analysis Model: Capabilities for Simulating a Waste Management System - 19131

The U.S. Department of Energy (DOE) Office of Nuclear Energy Office of Spent Fuel and Waste Disposition's Integrated Waste Management System (IWMS) program is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWMS are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The system architecture analyses consider alternative disposition pathways for commercial spent nuclear fuel (SNF) being stored dry in at-reactor Independent Spent Fuel Storage Installations (ISFSIs) or wet in spent-fuel pools. One method for evaluating these alternatives is simulating the movement of SNF throughout the system components. Successfully simulating a system on the scale of a national waste management system requires - accurate baseline information (e.g., historic fuel discharges from reactors and dry storage loadings at reactor sites), - modeling all logistic interactions between components of the proposed system, and - tracking and reporting all relevant details regarding the simulation's execution. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWMS. It is based on the Performance Analysis Tool, a simulation platform that has been used to model logistics for the Department of Defense and the Federal Emergency Management Agency. Using existing, government-owned software allows NGSAM developers to modify the software at the level of source code and to develop flexible usability and modeling solutions for developing and running simulations of the IWMS. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. NGSAM is designed with the flexibility to model and track the behavior of virtually any IWMS architecture assumptions or design considerations. For example, NGSAM can model multiple interim storage facilities or repositories or none at all. Similarly, all assumptions regarding shipment schedules, operating schedules, and reactor operation assumptions can be modeled and adjusted in NGSAM with relative ease. The development of NGSAM was targeted to waste management system analysts as the primary users. With that in mind, analysts from ORNL, Sandia National Laboratories (SNL), and Argonne National Laboratory (ANL) were intimately involved in the development process. That involvement extended beyond merely defining requirements. System analysts have been involved with design decisions and software benchmarking and tests, and they have provided subject-matter expertise where necessary. The result is a high degree of interaction between professional software developers and system analysts, with a multitude of benefits to both groups. This paper explores some of the underlying methodology of the agile software development process used in the creation of NGSAM. It illustrates the capabilities of NGSAM. Some example use cases are discussed as well, with relevant results providing context for how NGSAM is used by system analysts to gain a greater understanding of alternative waste management system architectures. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed-to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. (authors)

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French Geological Disposal Project: From Siting to Permitting - 19018

The future 'Cigeo' geological repository will be located in eastern France, at the border between the Meuse and Haute-Marne. Research has been carried out in this area since the 1990's, leading to the identification of a zone within the Paris basin whose geological characteristics make it suitable for the deep geological disposal of high- and intermediate-level long-lived radioactive waste, as well as a smaller zone of approximately 250 km{sup 2} (proposed in 2005) within which the characteristics of the host layer are very similar to those observed in the Underground Research Laboratory (URL). In 2010, based on additional surveys, a zone of approximately 30 km{sup 2} called ZIRA, in which the industrial underground facility would be constructed was defined. The repository structures within the ZIRA will be located at a depth of 525 metres. Half the surface area of the ZIRA is located in a wooded area, the other half being situated in grassland or farmland. All the subsequent design studies (preliminary and detailed) of Cigeo as well as the preparation of the licensing files (to be submitted in 2020) were conducted on the basis on this precise siting. (authors)

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