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DOE OSTI2020

Dual-Purpose Canister Filling Demonstration Project Progress Report

This report discusses the initial progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs) using filler materials to demonstrate that the probability of criticality in DPCs during disposal to be below the probability for inclusion in a repository performance assessment. In the initial phase of a multi-phase effort that will result in a full-scale demonstration, a computational fluid dynamics (CFD) model was developed to gauge the filling process and to uncover any unforeseen issues. The initial filling simulations of the lower region (mouse holes) of a prototypic DPC show successful removal of the inner space voids and smooth, even progression of the liquid level. In the initial phase, flow through a pipe that is similar to the drain pipe in a DPC will be investigated separately to gain valuable insight of flow regime inside a pipe. The initial experimental setups for validating the computational filling model have been designed, and the various assembly parts are being procured. The experience gained from the initial experiments will be applied to the next steps toward a full-scale demonstration and to the validation of multiphysics filling simulation models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Evaluation of Engineered Barrier Systems (Mid-year Status Report)

This document is a summary of the R&D activities associated with the Engineered Barrier Systems Work Package. Multiple facets of Engineered Barrier Systems (EBS) research were examined in the course of FY19 activities. This report is focused on delivering an update on the status and progress of modelling tools and experimental methods, both of which are essential to understanding and predicting long-term repository performance as part of the safety case. Specifically, the work described herein aims to improve understanding of EBS component evolution and interactions. Ultimately, the EBS Work Package is working towards producing process models for distinct processes that can either be incorporated into performance assessment (PA), or provide critical information for implementing better constraints on barrier performance. The main objective of this work is that the models being developed and refined will either be implemented directly into the Geologic Disposal Safety Assessment (GDSA) platform, or can otherwise be indirectly linked to the performance assessment by providing improved bounding conditions. In either the case, the expectation is that validated modelling tools will be developed that provide critical input to the safety case.

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

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

D and D Research on KM-IT Platform - 20494

The Knowledge Management Information Tool (KM-IT) is a web-based system developed to maintain and preserve the Department of Energy's (DoE's) knowledge base. The system was developed by Florida International University's Applied Research Center (FIU-ARC) with the support of the D and D community including the DOE Office of Environmental Management, and with the collaboration and support of the DoE's Energy Facility Contractors Group (EFCOG) and the former ALARA Centers at Hanford and Savannah. The KM-IT system is a community driven system tailored to serve the technical issues faced by the workforce across the DOE Complex. This KM-IT web-based interactive system is operational and available for use at www.dndkm.org and is currently comprised of the following modules: Web Crawler, Hotline, Technology, Hanford's ALARA Center Reports, Specialist Directory, Lessons Learned, Best Practices, Video/Picture Library, Vendors and Industry News. The Web Crawler dynamically searches through the KM-IT repository as well as the web and displays search results based on the search criteria. The Hotline allows registered users to post questions/problems related to D and D and to receive solutions from a subject-matter specialist. The Technology module serves as repository of D and D technologies and contains technology/ demonstration fact sheets and vendor information. D and D community members can add Lessons Learned, Best Practices, and Pictures/Videos to the D and D repository, after a formal approval process. The Vendor module provides a directory of commercial vendors who provide D and D related technologies, supplies, and services. KM-IT makes an excellent use of the knowledge that exists within the D and D community by allowing D and D project managers around the DOE complex to collaborate by sharing innovative ideas, past experiences, and practices and by maintaining a directory of subject matter specialists. FIU has developed the dedicated module on KM-IT platform to publish D and D research being performed at FIU. This module highlights current EM research efforts and activities in support of D and D being performed at FIU/SRNL associated with fixatives and intumescent products. FIU will expand this module to include the recent and current D and D research being performed across multiple DOE EM sites, national labs and universities. FIU will collaborate with SRNL, INL, ANL and other universities to explore and publish D and D research activities on KM-IT. As a result, the user community will have a centralized location to review new D and D related research across the DOE EM complex. (authors)

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DOE OSTI2017

A Bottom's-Up Model for Siting and Developing Consent-Based Disposal Solutions for SNF and HLW in the USA - 17332

At the end of 2016, the licensing and continued development of the USA's only candidate HLW-repository since 1987 at the Yucca Mountain (YM) site in Nevada had been on hold since 2010 pending the enactment of enabling legislation for one or more of the following paths forward: 1. Re-start of the YM HLW repository. 2. Consent-based siting and development (S and D) of new HLW-repositories for commingled disposal of commercially- (CHLW) and defense-generated (DHLW) HLW. 3. Consent-based S and D of a repository and/or deep boreholes for DHLW only. Public trust in the implementing and regulating entities and a definitive say in an apolitical S and D process by the directly affected parties (DAPs) are historical imperatives for gaining public acceptance and maintaining sustained support of any or all of them. However, due to the widespread negative legacy currently attributed to past politically-controlled S and D organizations, an even more politically controlled organizational structure and an amorphous, 'consent-based', S and D model like that introduced twice in proposed Senate bills since June 2013 will exacerbate, rather than mitigate, the main root causes to past contentions, lawsuits, delays, and cost-increases. We therefore recommend: 1) Assigning the responsibility for safe disposal of CHLW to an independent, chartered public corporation with requisite subject-matter expertise and adequate financial resources. 2) Enacting a related law that: a. Defines 'consent-based' in measurable terms; b. Assigns irrevocable veto rights to facility host state(s) based upon the majority opinion of its legal residents at pre-determined S and D milestones. c. Requires Congressional approval before the facility can open and close. (authors)

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DOE OSTI2020

Opportunities for an Integrated Web-Based Workbench for Data Access and Analysis - 20244

Management of environmental issues can require integration of multiple types of data and information, conducting data analysis and interpretation, and providing data visualization for effective communications. These data elements are important for site management to support regulator interactions and provide defensibility for remedial decisions. Databases and information repositories are core elements of managing data; however, efficient data access and analysis also enable effective site management. The U.S. Department of Energy (DOE) Hanford Site is an example of a complex site with a voluminous quantity of environmental data and a need for efficient site management. Different tiers of data and information tools have been developed and deployed to address site needs. These tools are configured for ready access via the web site interfaces and meet the rigorous quality requirements for environmental site management. Evolving efforts are focused on an integrated platform to meet site environmental management needs. In this platform, users can access site information at multiple levels of detail based on their need and permissions, so that data and associated analyses are presented within the context of the site mission and the user's management or technical needs. This concept is not only applicable at individual sites like Hanford but also applicable at other sites within the DOE complex. An integrated web-based architecture that links data visualization, data analytics, and management tools can provide holistic access to large data sets, minimize complexity, and maximize interactivity and technical communication. (authors)

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DOE OSTI2021

An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO 2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H 2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10 -7 year -1 ) is over estimated because it assumes that there is net UO 2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

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DOE OSTI2021

Dual-Purpose Canister Filling Demonstration Project Progress Report

This report discusses the progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs). Acceptable filler materials should demonstrate that the probability of criticality in DPCs during the disposal timeframe is below the probability threshold for inclusion in a repository performance assessment. This effort, which will ultimately result in a full-scale demonstration, includes computational fluid dynamics (CFD) modelling developed to gauge the filling process and to uncover any unforeseen issues. Filling simulations of the lower region (mouse holes) of a prototypic DPC show successfully simulate filling of the void space inside the canister and a smooth, even progression of the liquid level. Flow through a pipe that is similar to the drainpipe in a DPC is being investigated separately to gain valuable insight of the flow regime inside a pipe. Three physical experiments validating the computational filling model have been completed using surrogate liquids. One experiment using molten metal has been completed and the results demonstrate adequate filling of the canister void spaces. Although the scale experiment observed some void spaces related to shrinkage of the metal during cooling, the volume filled is expected to be sufficient to meet the purpose of moderator exclusion. Further experiments will scale up the geometry, moving towards a full-sized DPC demonstration, and provide a more rigorous investigation of candidate filler materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Drop Analysis of Department of Energy Standard Canister with Fort Saint Vrain SNF

DOE manages over 300 types of SNF, most of which are located at the INL site. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF at the INL site for ?road-ready dry storage?. ?Road-ready dry storage? is a SNF management concept where SNF is packaged into dry and sealed canisters, which are then placed in on-site storage in anticipation of later transportation. The forward-looking goal of the Demonstration is establishing the foundation for a large-scale road-ready dry storage program at the INL site. The Demonstration will first package Fort Saint Vrain SNF currently stored at INL into several DOE Standard Canisters. These Standard Canisters will then be loaded into another commercial transportation or storage containment system (e.g., storage cask or transportation package). The Standard Canister is a class of standardized canisters designed for containing the large variety of DOE-managed SNF during interim storage, transportation and/or disposal at a geological repository. One critical aspect of road-ready dry storage is the ability to license the DOE Standard Canisters and its associated transportation package to 10 CFR 71. Depending on the SNF and transportation strategy, the Standard Canisters may have to maintain structural integrity under normal conditions of transport and hypothetical accident scenarios (i.e., drop events). The Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs. However, no analysis has been completed to support the recent Demonstration. This analysis will consider the Ø0.5 m × 5.1 m Standard Canister under drop scenario(s) considered in previous INL tests and analyses, including the 9 m drop at 80 degree off vertical. However, this analysis will consider the more recent Fort Saint Vrain loading configurations proposed for the Demonstration. This analysis will performed using strain-based acceptance criteria established by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3. It will be compared to previous analyses and form the foundation of further formal calculations that will be used to support licensing efforts of the road-ready dry storage system at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Modeling of Radiolytic Production of Oxidants Relevant to Corrosion of a Used Fuel Container in Deep Geologic Repository Environments - 17641

The Canadian used nuclear fuel disposal concept involves a deep geological repository using a multi-barrier system. A key barrier is the used fuel container (UFC). It consists of an inner vessel of carbon steel, which provides the structural strength to withstand repository loads, and an outer coating of copper acting as an external corrosion barrier. Relative to other design concepts, coated containers have thinner walls that provide less shielding for γ-radiation, raising the possibility of radiolysis-induced corrosion on the exterior of the container. Also, any trapped aerated water inside the container or in the bentonite clay may condense on the container surface, be exposed to γ-radiation and influence the container corrosion. Understanding the effect of radiation on container corrosion is one key factor to consider in the assessment of the integrity and longevity of the UFC. Humid air radiolysis produces HNO{sub 3} as the dominant oxidizing species. A conservative hemispherical water droplet model was developed to predict the accumulation of HNO{sub 3} and its impact on corrosion. Radiolysis of groundwater in the presence of Cl{sup -} mainly produces H{sub 2}O{sub 2} and HOCl. The concentration of these species is <10 mM and their sum remains almost constant. Radiation exposure tests showed radiation-induced corrosion on carbon steel and Cu. A systematic study of the Cu corrosion and the corrosion rate as a function of [H{sub 2}O{sub 2}], [HOCl], pH and γ-radiation is on-going. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

An international study on THM modelling of the full-scale heater experiment at Mont Terri laboratory

We present results from an international model comparison study of the Full-Scale Emplacement (FE) experiment in Opalinus Clay at the Mont Terri Laboratory, Switzerland. Based on a provided parameter set the teams decided which parameters they adopted for their models, whether they considered the excavation and the ventilation phase in addition to the heating phase and if they included technical features like the shotcrete or the EDZ. The teams were able to reproduce the measured parameters temperature, relative humidity and pore pressure. The modelled results for temperature agree very closely between the teams especially in the sensors in Opalinus Clay. All teams were able to reproduce the redistribution of water in the bentonite backfill due to heating. The evolution of the relative humidity showed similar trends with differences in the intensity of the dry out effect. To model the pore pressure evolution is more complex because it comprises the full interaction of the coupled THM processes. The spread between the pore pressure modelled by the teams was larger, with some teams overestimating the pressure increase due to heating and some teams overestimating the extent of drainage. The agreement of modelled results with measurements improves with larger distance to the heater. We conclude that the EDZ and the shotcrete potentially influence the behaviour of the rock causing higher differences closer to the heater. Further research is needed to better implement those influences into the models. Based on the calibrated models, the future evolution of temperature, relative humidity and pore pressure was predicted over the next 10 years following a change of the heat power applied in 2023 and 2024. Again, the predicted temperatures agree very closely between the teams. Most teams do not expect an increase in relative humidity during the next 10 years after the initial dry-out.

58 GEOSCIENCES↗
DOE OSTI2018

KM-IT Mobile Platform for D and D - 18300

Knowledge management (KM) is the practice or process responsible for gathering, analyzing, storing and sharing insights, experiences, knowledge, and information within an organization or community. The main focus of the Knowledge Management Information Tool (KM-IT) for the deactivation and decommissioning (D and D) community is to enhance safety through improved efficiencies (e.g., by reducing the need to rediscover the knowledge and experience gained over time and to promote the reuse of the existing knowledge). KM-IT serves as a centralized information repository and a common interface for D and D related activities. It facilitates the gathering, analysis, storage and sharing of knowledge and information within the D and D community. FIU has developed eight active modules as part of the web-based mobile application for the KM-IT, which can be accessed from m.dndkm.org: Vendor, Specialist Directory, Technology, Picture Library, Hotline, Lessons Learned, Best Practices and Fixatives. All of the mobile modules are scaled-down versions of the desktop module available at www.dndkm.org and publish data from the KM-IT database to the D and D community. FIU has recently developed a native mobile application for Android mobile devices for selecting D and D fixatives. This application runs natively on mobile devices and displays data through the fixative services which uses REST-based API for communication across mobile platforms. Xamarin Forms are used to develop the front-end application for the Android operating system (OS) and will be used for future platforms (iOS and Windows) as well. The native fixative application can be installed on mobile devices to search information related to fixatives, decontamination gels and strippable coatings applicable to D and D. The KM-IT platform includes a comprehensive database of commercially available fixatives and other contamination control products. Search functionality allows the user to filter and sort these products according to the search criteria or keyword. Three levels of refinement are available via a drill down menu for the criteria. (authors)

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

Criticality effect of long-term changes in material compositions and geometry in disposal canisters

For the Swedish final repository the basic requirement concerning margin to criticality is that k{sub eff} should be less than 0.95. This criterion is always fulfilled if the disposal canister is dry. If it is assumed that a canister is filled with water it is shown that the criticality criterion is fulfilled using burnup-credit for PWR fuel and credit for burnable poison for BWR fuel. If water is assumed to leak in, long term chemical processes will change the material compositions and geometries in the canister during the one million year analysis period. The main process expected to affect significantly the chemical composition of the materials in the canister is the corrosion of cast iron and carbon steel of the insert, the corrosion of the stainless steel, Inconel or Zircaloy in the fuel element structural material, the corrosion of Zircaloy cladding and finally the corrosion of spent fuel itself. During metal corrosion, various anoxic corrosion products are formed, usually with lower density than the metal itself, causing thus a volume increase in the canister, accompanied by geometrical changes. In this paper the reactivity effects of these changes are analysed and results presented. It is shown that the largest reactivity increase is due to the decrease of the water gap between the fuel and the fuel channel, due to the formation of Magnetite. Other process that possibly will increase the reactivity is if Siderite is produced besides from Magnetite, the production of Hydrogen, corrosion of the fuel and radial movements of fuel rods. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Status Update on the High Precision Isotopic Measurements on High Burnup LWR Fuel in 2020

The US Department of Energy (DOE) Office of Nuclear Energy (NE) is currently investigating the feasibility of directly disposing dual-purpose (storage and transportation) canisters (DPCs) in a spent nuclear fuel (SNF) repository. Criticality during the repository performance period (10,000 years or more) is one of the major concerns related to direct disposal of DPCs, specifically as the system undergoes degradation in the repository environment and timeframe. Oak Ridge National Laboratory (ORNL) is developing an as-loaded criticality analysis methodology using full (actinides + fission products) burnup credit that exploits the inherent criticality margin associated with actual canister-specific loading configuration. Burnup credit criticality analysis requires validation of the depletion/decay codes used to generate the burned isotopic inventory of an assembly by comparing the code-predicted inventory with experimentally determined isotopic data. Currently, isotopic measurement data for boiling water reactors (BWR) SNF are limited, and additional measurements will be highly beneficial for BWR burnup credit analysis, which is essential to demonstrate disposability of BWR DPCs. Moreover, new pressurized water reactor (PWR) samples of isotopic measurements will expand the PWR sample population and consequently will greatly improve sample statistics. In turn, this will reduce uncertainty in the computational determination of the isotopic composition of commercial SNF and eliminate additional penalties currently used for lack of data. As part of the isotopic analysis task, eight diverse, high burnup (HBU) samples from several PWR rods are being dissolved and isotopically analyzed to provide high-quality measurement data for PWR SNF to reducing the uncertainties associated with PWR isotopic depletion validation. The isotopics of interest focus on nuclides important to burnup credit, shielding, and decay heat in PWR SNF. Additionally, eight high burnup BWR fuel samples from the Limerick nuclear power plant have been identified for measurement to support reducing uncertainties associated with BWR isotopic depletion validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Influences of Mg Ions and Al Ions on Deposition Behavior of Supersaturated Silicic Acid - 20261

Silicic acid can change into various species such as deposit, colloidal, and soluble form under certain conditions: for example, gradual pH changes around geological disposal repositories. In particular, silicic acid deposits narrow groundwater flow paths, thereby restricting the flow of groundwater. Such effects may contribute to a delay in radionuclide migration. Previous studies have reported that silicic acid supersaturated in sodium and calcium ions can narrow the flow paths. This study examined the effect of magnesium (Mg) and aluminum (Al) ions on the deposition of supersaturated silicic acid on the solid phase. The deposition behavior of supersaturated silicic acid on the solid phase was investigated in a batch system while changing the initial supersaturation concentration, the coexisting ion concentration, and the amount of added solid phase. In the presence of Mg ions, silicic acid precipitation increased with increasing Mg ion concentration due to the influence of the electrolyte, and crosslinking reactions. The coexistence of Al ions significantly suppressed silicic acid deposition due to the formation of a covering on the solid surface during Al ion hydrolysis under conditions of high Al concentration (1.4 mM). On the basis of these experimental results, the apparent deposition rate constant of silicic acid was estimated to be in the range from 10{sup -11} to 10{sup -10} m/s. The Damkoehler number (the dimensionless number expressed by the ratio of the characteristic time of mass transfer and the chemical reaction constant) was sufficient to cause the clogging of the host rock by deposition. This suggests that supersaturated silicic acid deposition, regardless of the presence of Mg ions (and Al ions), contributes to improving barrier performance against radionuclide migration in the host rock. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage

The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times.

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

Exploring the Use of Muon Momentum for Detection of Nuclear Material Within Shielded Spent Nuclear Fuel Dry Casks

The inspection of the storage containers that house used nuclear fuel is an important issue facing the nuclear industry; currently, there are limited options available to provide for even minimal inspections. During the long storage life of the fuel, intermediate handling and transportation activities may take place that can create concerns regarding fuel integrity (safety) and continuity of knowledge (safeguards and security). Accidents and natural disasters also contribute to this concern. The reliable and safe management of used nuclear fuel is among the critical tasks to be addressed for the advancement of fission-based nuclear energy in the United States and elsewhere. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of used nuclear fuel have been generated in the United States. The postponement of the opening of a permanent repository until 2048 will lead to increased used fuel accumulation at reactor sites. A rapid increase in storage installations with more than 9,000 casks storing {approx}420,000 assemblies is anticipated by 2050 even if no new nuclear power plants are built. After used nuclear fuel has been placed inside a cask, the cask is sealed, not allowing for visual inspection. Conventional methods, e.g., x-rays, are limited in that they cannot penetrate dense, well-shielded objects. More sophisticated techniques, such as penetrating neutrons or proton radiography, require the use of an expensive accelerator. Efforts by Ziock et al. using neutron and gamma-ray imagers concluded that '...the signatures are deemed insufficient to uniquely identify individual casks of the same type...'. It was found that the gamma-ray and neutron images were dominated by diffusion and thermalization effects resulting from the large amount of shielding. One promising mode of inspection would use cosmic-ray muons. These muons are naturally generated, are able to penetrate dense materials, and are incident from all directions above the horizon. Tomography performed using these muons could potentially provide efficient, inexpensive, remote assessment of used nuclear fuel. However, current muon applications are limited in that they have not included momentum measurement. If momentum measurement could be incorporated, that would represent a big step forward because it would allow more accurate estimation of the muon observables, e.g., scattering variance. This approach has the potential to reconstruct the contents of a cask with improved resolution and identify partial defects, fuel assemblies, and individual fuel rods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Corrosion Dynamics of Carbon Steel in Small, Stagnant Volume under Radiation

The long term disposal of used nuclear fuel involves a deep geological repository (DGR) with the key barrier being the Used Fuel Container (UFC) consisting of a copper-coated carbon steel vessel. NWMO's Mark II design UFC is assembled by welding the semi-spherical head to a cylindrical body. Crevice is formed between the two parts along with the weld region. The objective of this research is to develop and validate a reliable corrosion model for long-term corrosion of CS under DGR condition. Water radiolysis produces chemically reactive species that affect corrosion rates and pathways. This work explores the synergistic effects of γ-radiation and solution properties on the dynamics of carbon steel corrosion. In Stage 1, corrosion rate is controlled by dissolution of iron to the solution phase. In Stage 2, the controlling processes are the growth of iron hydroxide gel. In Stage 3, the main iron redox reaction is the conversion between different iron oxides, and the overall corrosion rate is negligible. The concentration of Fe{sup 2+} (aq) near the metal surface reaches its solubility limit very quickly in a small, stagnant volume. Diffusion from crevice to bulk solution is more significant with larger crevice size. Porous, granular form of magnetite loosely attached to the metal surface is formed from Fe{sup 3+} in solution via reduction by H{sub 2}O{sub 2}. As a source of continuous formation of H{sub 2}O{sub 2}, radiation accelerates oxidation (Fe0 → Fe{sup 2+} → Fe{sup 3+}). Once the oxidation is almost complete, H{sub 2}O{sub 2} can act as a reducing reagent, reducing Fe{sup 3+} species back to Fe{sup 2+} forming more stable mixed Fe{sup 2+,3+} species (magnetite and maghemite)

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