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At least 865 records · Page 48

Panel Session 14: US DOE WIPP - Future Plans for the Utilization of a National Asset

This panel focused on the next phase of restoring the Waste Isolation Pilot Plant (WIPP: located in Carlsbad, New Mexico, USA) to full disposal operations. The US DOE Carlsbad Field Office (CBFO) is funded for two Capital Asset Projects that will facilitate the restoration of WIPP to full capability and allow for future repository space planning. The panel will include updates on the capital projects, infrastructure improvements to extend WIPP's operational life, regulatory actions in support of capital projects, and improvements in new processes to implement WAC and DSA enhancements. This panel is filled by CBFO and WIPP contractor leaders to provide the big picture leading to WIPP's future. Presentations proposed include: 1 - 2018 WIPP Year in Review (Bruce Covert, Nuclear Waste Partnership, LLC (USA)); 2 - TRU Waste Shipments and WIPP Operations (Mark Pearcy, Nuclear Waste Partnership, LLC (USA)); 3 - Projects and Activities at WIPP (Ronald Gill, US DOE (USA)); 4 - The Future of WIPP (J.R. Stroble, US DOE (USA), Todd Shrader, US DOE (USA))

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The Proposed Operations Master Plan of Cigeo - 17013

The proposed operations master plan of Cigeo describes the so-called 'reference progression' of the Cigeo project as envisaged by Andra at the end of the preliminary design studies. It aims at clarifying the objectives of the industrial pilot phase and present the choices offered by the reversibility for the management of the Cigeo project. The reference progression includes the inventory of the waste to be disposed of in Cigeo used for the dimensioning (73 600 m{sup 3} of ILW-LL and 10 100 m{sup 3} of HLW). Cigeo is designed to store the waste already produced and those that will be produced in the future by the existing nuclear installations and those whose creation was authorized, until the predictable end of their operation and their dismantling. All spent fuel is assumed to be reprocessed. The reference progression also includes the duration and the projected milestones of the construction and operation phases of Cigeo until termination. Across the operating life of Cigeo, the extensions of the Cigeo repository areas proceed by successive incremental phases, from the start of the industrial pilot phase (2025) to the final closure operations planned by 2150. Reversibility is defined as the ability to offer to the next generations different choices in terms of radioactive waste long-term management. The implementation of the reversibility principle is based on both governance and technical project management tools. (authors)

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Radiological Characterisation of VLLW Produced at High-Energy Particle Accelerators - 19118

In this paper we describe the characterisation method of very-low-level radioactive waste (VLLW) produced at the European Organization for Nuclear Research (CERN). The method is based on the calculation of a preliminary radionuclide inventory using either analytical or Monte Carlo codes, on the measurement of γ-emitters and on the quantification of pure-α, pure-β and low-energy X-ray emitters, called difficult-to-measure (DTM) radionuclides, using the so-called scaling factor (SF) technique. The first stage of the characterisation process is the calculation of the radionuclide inventory, which is a list of radionuclides that can potentially be produced via the interaction of primary and secondary particles with the surrounding media. Once the preliminary radionuclide inventory is obtained, the γ-emitting radionuclides are measured via γ-ray spectrometry on each package of the waste population. The major γemitter, called key nuclide (KN), is also identified. The scaling factor method estimates the activity of DTM radionuclides by checking for a consistent and repeated relationship between the key nuclide and the activity of the difficult to measure radionuclides from samples. Between 2016 and 2018, several hundred samples were collected at CERN and over 1000 radiochemical analyses were performed to estimate robust statistics to quantify scaling factors for the most common DTM radionuclides, such as H-3, Fe-55 and Ni-63. This paper illustrates the sample collection strategy, the tests performed to choose the most appropriate statistical models describing the behaviour of scaling factors and a complete quantification of the uncertainties associated with the characterisation process. Finally, we present the results obtained during the characterisation of over 2000 m{sup 3} radioactive metals and cables disposed of at the French National repository for VLLW. (authors)

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Comparative analysis of uranium bioassociation with halophilic bacteria and archaea

Rock salt represents a potential host rock formation for the final disposal of radioactive waste. The interactions between indigenous microorganisms and radionuclides, e.g. uranium, need to be investigated to better predict the influence of microorganisms on the safety assessment of the repository. Hence, the association process of uranium with two microorganisms isolated from rock salt was comparatively studied. Brachybacterium sp. G1, which was isolated from the German salt dome Gorleben, and Halobacterium noricense DSM15987T , were selected as examples of a moderately halophilic bacterium and an extremely halophilic archaeon, respectively. The microorganisms exhibited completely different association behaviors with uranium. While a pure biosorption process took place with Brachybacterium sp. G1 cells, a multistage association process occurred with the archaeon. In addition to batch experiments, in situ attenuated total reflection Fourier-transform infrared spectroscopy was applied to characterize the U(VI) interaction process. Biosorption was identified as the dominating process for Brachybacterium sp. G1 with this method. Carboxylic functionalities are the dominant interacting groups for the bacterium, whereas phosphoryl groups are also involved in U(VI) association by the archaeon H. noricense.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE SFWST Campaign R&D Roadmap Update Rev.1.

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Departmentof Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel and Waste Disposition(SFWD) is conducting research and development (R&D) on deep geologic disposal of spentnuclear fuel (SNF) and high-level nuclear waste (HLW). R&D addressing the disposal ofSNF/HLW in the U.S. is currently generic (i.e., "non-site-specific") in scope, following thesuspension of the Yucca Mountain Repository Project in 2010. However, to prepare for theeventuality of a repository siting process, the former Used Fuel Disposition Campaign (UFDC) ofDOE-NE, which was succeeded by the SFWST Campaign, formulated an R&D Roadmap in 2012outlining generic R&D activities and their priorities appropriate for developing safety cases andassociated performance assessment (PA) models for generic deep geologic repositories in severalpotential host-rock environments in the contiguous United States. This 2012 UFDC Roadmap alsoidentified the importance of re-evaluating priorities in future years as knowledge is gained fromthe DOE's ongoing R&D activities.

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The iodine-129 paradox in nuclear waste management strategies

Nuclear energy has an important role in the low-carbon energy transition, but the safety of spent nuclear fuel (SNF) management remains a public concern. Here, in this study, we investigate the interplay between waste management strategies and their environmental impacts with a particular focus on a highly mobile and persistent radionuclide, iodine-129 (I-129), which is the dominant risk contributor from SNF disposal and at existing groundwater contamination sites. The results show that the current recycling practice releases more than 90% of I-129 in SNF into the present-day biosphere using an isotropic dilution strategy, whereas the direct disposal of SNF in geological repositories is likely to delay and reduce the release by 8 orders of magnitude. In addition, our data synthesis of surface water concentrations near four nuclear facilities shows that the release-dilution strategy results in lower concentrations than regulatory standards, while insufficient waste isolation in the past has resulted in locally high concentrations within one site. Our analysis suggests that it is essential to consider effluents more explicitly as a part of the waste, that as society moves from dilution to isolation of waste, the potential risks of waste isolation to local regions should be carefully evaluated, and that excessive burdens of proof could hinder or discourage waste isolation. Comprehensive waste management strategies—considering not just volume but also mobility, isolation technologies and ultimate fates—are needed for persistent contaminants. This study offers valuable insights for optimizing the management of SNF and other persistent contaminants.

Wainwright, Haruko M. [Massachusetts Inst. of Tech↗

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

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Analysis of the Needs for Long Term Memory and Knowledge Preservation Relating to Radioactive Waste Disposal Facilities - 17213

Proposals for a set of provisions for long-term memory and knowledge to be defined for a radioactive waste disposal facility, near surface or deep underground, address two primary motives, related to two ethical principles. The first motive is to prevent future generations from interfering involuntarily with the repository. This requires maintaining awareness of the repository, and addresses the ethical principle of protection of man and environment. The second motive is to provide future generations all the available relevant information which might help them make informed decisions about intentional actions, and assess the consequences. This requires transmitting detailed knowledge of the repository, and addresses the ethical principle of preservation of freedom of action. The set of provisions to be implemented with respect to each of these motives may not be the same. In order to define and assess the set of provisions, it is also useful to identify the various components of the process of transmission of a given message, or set of messages, to future generations. Three sub-processes have been identified: (i) 'memorization', at the producer stage, where a full set of information to be transmitted is identified, organized and expressed; (ii) 'preservation', where the potential durability of records is extended, the preservation conditions are controlled and where the records may be restored, if their status is degraded; (iii) 'access', at the receiver stage, where the receiver has to be notified of the existence of the information, to find it and to interpret it properly. As a failure of transmission to future generations would result from the failure of any of the subprocesses, a minimal set of provisions may be defined from this decomposition. (authors)

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Effects of Borate and Organics on U(VI) Solubility in WIPP Brine

The solubility of uranium (VI) in Waste Isolation Pilot Plant (WIPP)-relevant brine was determined to support ongoing WIPP recertification activities. This research was performed by the Los Alamos National Laboratory Carlsbad Operations (LANL-CO) Actinide Chemistry and Repository Science Program (ACRSP). The WIPP Actinide Source Term Program (ASTP) did not develop a model for the solubility of actinides in the VI oxidation state. The solubility of UO 2 2+ , in the absence of WIPP specific data, is presently set to be equal to a conservatively high 1 mM within the WIPP Performance Assessment (PA) for all expected WIPP conditions (SOTERM, 2019) as selected at the recommendation of the Environment Protection Agency (EPA) (EPA, 2005). According to the current WIPP chemistry model assumptions and conditions, the expected pC H+ is about 9.5 and controlled by MgO buffering CO 3 2- . The goal of this study is to perform screening experiments that account for the contributions of organics and borate on uranium solubility. In this report, the solubility of U(VI) was determined at pC H+ 9 WIPP brine in the absence or presence of borate and organics at under-saturation approach. Experiments were equilibrated for about 135 days. Organic compounds present in WIPP waste can form strong complexes with actinides and can affect the oxidation states of actinides. The organic compounds addressed in WIPP performance assessment include EDTA (Ethylenediaminetetraacetic Acid), oxalate, citrate, and acetate (SOTERM, 2019). These data quantify the effects of WIPP-relevant concentrations of borate and organics effects on the solubility of U(VI) to challenge the predictions of the WIPP actinide model and inform decisions and recommendations made in the upcoming recertification of the WIPP (CRA-2024). The experiments performed were done according to the U.S. Department of Energy (DOE) approved Test Plan entitled “Experimental Strategy to Challenge Actinide Solubility Predictions” and designated LCO-ACP-26. All data reported were obtained under the LANL-CO Quality Assurance Program, which is compliant with the DOE Carlsbad Field Office, Quality Assurance Program Document (CBFO/QAPD) (QAPD, 2017).

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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.

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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

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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.

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