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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Gap Analysis to Guide DOE R&D in Supporting Extended Storage and Transportation of Spent Nuclear Fuel: An FY2019 Assessment (Final Report)

This report is a condensed version of previous reports identifying technical gaps that, if addressed, could be used to ensure the continued safe storage of SNF for extended periods and support licensing activities. This report includes updated gap priority assessments because the previous gap priorities were based on R&D performed through 2017. Much important work has been done since 2017 that requires a change in a few of the priority rankings to better focus the near-term R&D program. Background material, regulatory positions, operational and inventory status, and prioritization schemes are discussed in detail in Hanson et al. (2012) and Hanson and Alsaed (2019) and are not repeated in this report. One exception is an overview of the prioritization criteria for reference. This is meant to give the reader an appreciation of the framework for prioritization of the identified gaps. A complete discussion of the prioritization scheme is provided in Hanson and Alsaed (2019).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

International Collaboration Activities in Geologic Disposal Research (FY24 Progress Report)

This report describes the FY24 status of international collaboration regarding geologic disposal research in the Office of Spent Fuel and Waste Science and Technology (SFWST) in the Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of SFWST is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal.

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

International Collaboration Activities in Geologic Disposal R&D: Spent Fuel and High-Level Waste Disposition (FY25 Progress Report)

This report describes the FY25 status of international collaboration on geologic disposal research and development (R&D) in the Office of Spent Fuel and High-Level Waste Disposition (SFHLWD) in the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of this office is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal. R&D in geologic disposal is administered in SFHLWD’s Disposal R&D Campaign.

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

Vibrational metrics for the evaluation of internal conditions in a scaled nuclear dry storage cask

In many countries, permanent repositories for high burnup spent nuclear fuel assemblies have yet to be established. As a result, much of the spent fuel from nuclear reactors is contained within dry storage casks, with many being beyond their designed service life. The assessment of the internal structural integrity of these casks and the fuel assemblies contained therein is of critical importance for both extended storage and transport to permanent repositories. The large size, structural complexity, and inaccessibility of the interior of the casks make this task challenging. To address these difficulties, a 1:6 scaled model based closely on the design of a Transnuclear (TN-32) dry storage cask was fabricated to facilitate controlled studies of these structures in the laboratory. Vibrational spectroscopy was used to evaluate the state of the cargo and internal structures within the cask utilizing only measurements on the outer surface. Using modes identified through Finite Element modeling corresponding to those previously measured on a full-scale TN-32 cask, we report on the development of amplitude- and phase-based metrics that are sensitive to internal conditions in the lab cask. Steel rod bundles and steel shot were used as surrogates for intact and damaged fuel assemblies, and various internal configurations of these materials were investigated. Finally, the metrics were based on acquired spectra involving the (1, 2) global bending mode and the (2, 1) radial-with-shearing mode. The results show that the metrics are sensitive to the condition of a single assembly and have some ability to determine the locations of damaged and empty slots.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2024

Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods

At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Preliminary Modeling of Chloride Deposition on Spent Nuclear Fuel Canisters in Dry Storage Relevant to Stress Corrosion Cracking

This study presents a method that can be used to gain information relevant to determining the corrosion risk for spent nuclear fuel (SNF) canisters during extended dry storage. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. This study uses well-developed computational fluid dynamics and particle-tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. The deposition rate is determined for a vertical canister system and a horizontal canister system, at various decay heat rates with a uniform particle size distribution, ranging from 0.25 to 25 µm, used as an input. In all cases, most of the dust entering the overpack passed through without depositing. Most of what was retained in the overpack was deposited on overpack surfaces (e.g., inlet and outlet vents); only a small fraction was deposited on the canister itself. These results are provided for generalized canister systems with a generalized input; as such, this paper is intended to demonstrate the technique. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Nuclear Energy Office of Spent Fuel Waste Science and Technology, which is tasked with doing research relevant to developing a sound technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Effects of Residual Water on Storage Canister Internal Components

This report describes an evaluation of the spent nuclear fuel cladding and canister internal materials’ corrosion/oxidation due to (inadvertent) residual free water inside a dry storage canister post-dryout. Drying of spent nuclear fuel (SNF) and the impact of (inadvertent) residual water in a SNF canister is being addressed under the DOE-NE Spent Fuel and Waste Science and Technology (SFWST) program to ensure the safe extended dry storage and transportation of commercial SNF. A long-standing understanding has been that the amount of residual water in a storage canister after vacuum drying is not expected to be much more than trace amounts of 0.43 gram mole. However, recent findings from the High Burnup Demonstration project, and an Integrated Research Project show that residual free water well above the amount of approximately 0.43 gm-moles that had been assumed for a 3 torr rebound pressure, may remain within an SNF canister following prototypic drying. In fact, an NRC-NMSS sponsored study considered residual water amount as high as 55 moles. Considering the recent findings and the NRC-sponsored work, this present work was conducted to study the effects of unspecified amount of residual water on oxidation and corrosion of canister internals, i.e., it was assumed that the residual water amount is not limiting. It was assumed that the spent nuclear fuel content in the canister is not exposed to the internal environment, i.e., the packaged canister does not contain any damaged (breached) fuel.

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

Cosmogenic nuclide and solute flux data from central Cuban rivers emphasize the importance of both physical and chemical mass loss from tropical landscapes

We use 25 new measurements of in situ produced cosmogenic 26 Al and 10 Be in river sand, paired with estimates of dissolved load flux in river water, to characterize the processes and pace of landscape change in central Cuba. Long-term erosion rates inferred from 10 Be concentrations in quartz extracted from central Cuban river sand range from 3.4–189 Mg km –2 yr –1 (mean 59, median 45). Dissolved loads (10–176 Mg km –2 yr –1 ; mean 92, median 97), calculated from stream solute concentrations and modeled runoff, exceed measured cosmogenic- 10 Be-derived erosion rates in 18 of 23 basins. This disparity mandates that in this environment landscape-scale mass loss is not fully represented by the cosmogenic nuclide measurements. The 26 Al / 10 Be ratios are lower than expected for steadystate exposure or erosion in 16 of 24 samples. Depressed 26 Al/ 10 Be ratios occur in many of the basins that have the greatest disparity between dissolved loads (high) and erosion rates inferred from cosmogenic nuclide concentrations (low). Depressed 26 Al/ 10 Be ratios are consistent with the presence of a deep, mixed, regolith layer providing extended storage times on slopes and/or burial and extended storage during fluvial transport. River water chemical analyses indicate that many basins with lower 26 Al/ 10 Be ratios and high 10 Be concentrations are underlain at least in part by evaporitic rocks that rapidly dissolve. Our data show that when assessing mass loss in humid tropical landscapes, accounting for the contribution of rock dissolution at depth is particularly important. In such warm, wet climates, mineral dissolution can occur many meters below the surface, beyond the penetration depth of most cosmic rays and thus the production of most cosmogenic nuclides. Our data suggest the importance of estimating solute fluxes and measuring paired cosmogenic nuclides to better understand the processes and rates of mass transfer at a basin scale.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water Reactor Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages

There is an increased interest in operating commercial light-water reactors (LWRs) in the United States with improved economics that would result from longer fuel cycle lengths, fewer and shorter refueling outages, and fewer fuel assemblies requiring storage at the back end of the fuel cycle. To support this, fuel discharge burnups, as well as initial 235 U enrichments, must be higher than those used in current commercial LWRs. The typical upper limit considered for assembly average burnup in this report is 75 gigawatt-days (GWd) per metric ton of uranium (MTU), as opposed to the current typical upper bound of approximately 62 GWd/MTU. The upper limit considered for initial 235 U enrichment is 8 weight percent (8 wt %), as opposed to the current regulatory limit of 5 wt %. The enrichment range from 5 to 8 wt % is referred to in this report as extended enrichment . To investigate the effect of high burnup and extended enrichment conditions on dose rates and burnup credit for dry storage casks and transportation packages, a fuel assembly and irradiation parametric study was performed. The conclusions from this study will assist U.S. Nuclear Regulatory Commission staff in reviewing applications for dry storage casks and transportation packages that contain high-burnup and extended enrichment fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project

The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2022

Ageing Management for Extended Long-Term Dry Storage of Spent Nuclear Fuel and Transportation

The principal objectives of the Coordinated Research Project (CRP) and CRADA were to (1) investigate how ageing effects leading to degradation of materials used in the spent fuel dry storage systems could be managed by ageing management programs (AMPs) and (2) using existing AMPs as a basis, establish guidance on how to develop, generate, and maintain AMPs for dry storage systems of spent nuclear fuel (SNF) that can be accomplished in various ways. The PI (Dr. Liu) is an internationally recognized expert on ageing management for license renewal of nuclear power plants and independent spent fuel dry storage installations. Dr. Liu was invited by IAEA to participate in this CRP and served as Chair of the Working Group, supporting the IAEA CRP Lead, with other members from Argentina, the Czech Republic, France, Germany, Hungary, Japan, Pakistan, Spain, Switzerland, United Kingdom, and the United States of America. Argonne’s scope of work in the CRADA for the CRP included (1) ageing management guidance documents developed by Argonne for DOE and used by the Nuclear Regulatory Commission (NRC) and industry and (2) research leveraged from the DOE Office of Nuclear Energy on the mechanical properties of high-burnup fuel cladding and the ARG-US remote monitoring systems technology developed for the DOE Packaging Certification Program, Office of Packaging and Transportation, Office of Environmental Management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

LightMAT - Continuous fiber malleable thermoset composites with sub-1-minute dwell times; validation of impact performance and evaluation of the efficacy of the compression forming process (CRADA 409 Final Report)

In spite of carbon fiber composite materials’ significant impact on light-weighting, performance, and efficiency in the aerospace industry, adoption of these materials has been slow in high-volume industries such as automotive. Two major barriers to adoption are cost and cycle-time. Current approaches to thermoset composite part production are dependent on in-mold curing of thermosets which are not optimal for high volume, high throughput production due to limitations and inefficiencies associated with the in-mold cure itself. In addition, ancillary activities, such as transport and storage of shelf-life-limited uncured, or B-staged prepreg materials typically necessitate refrigerated transport and storage which significantly impacts the economic and energy costs associated with manufacturing. Furthermore, limited out-life means that trimmings cannot typically be used as they partially cure during initial production leading to high scrap rates (+20% is common). Finally, the need to cure in-mold drives cycle times to multiple minutes in the best cases, and 10’s of minutes to hours in most cases. Mallinda is developing polyimine malleable thermoset prepreg composite materials which have excellent mechanical properties (100 GPa tensile modulus, 2 GPA tensile strength, 2.4% elongation at break) and high operating temperatures (Tg>200°C). At scale, polyimine resins are commensurate in price with commodity epoxy resins. What distinguishes malleable thermoset prepreg from traditional thermoset prepreg materials, is that they are fully cured during Mallinda’s roll-to-roll production of prepreg laminate. This results in 5 key value-differentiating benefits. First, it simplifies manufacturing logistics by enabling ambient transportation and storage, and by significantly extending out-life and shelf-life almost indefinitely. Second, elimination of autoclave curing reduces the economic and energy costs to the customer. Third, scrap rates can be reduced as malleable thermoset prepreg materials are directly reusable. Fourth, the manufacturing consolidation step can be roughly 10x faster than traditional thermosets, because the resin is already cured. Parts can be made via compression forming by the application of heat and pressure to quickly vitrify and consolidate a multilayer part – easily leading to sub 3-minute cycle times (at lab scale we have demonstrated a 20 second dwell time, with room for further optimization). Finally, the closed-loop cradle-to-cradle solution-based recyclability of malleable thermoset composites can also contribute significantly to the future of sustainable lightweight materials. The focus of this project was the development, optimization and validation of malleable thermoset composite materials which exhibit manufacturing cycle times of 3-minute or less, high speed impact performance on par with incumbent technologies, and defect-free consolidation of 3 dimensional parts. Government funding is required on this project because Mallinda is a start-up company whose focus is the development and scale-up of the described technology for introduction to the transportation segment. Unlike incumbent industry players, Mallinda is bringing a completely new and highly relevant malleable thermoset technology to bear on the problem of high throughput composites. As a small business, Mallinda relies on a combination of government R&D funding and private capital to perform development work.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Continuous fiber malleable thermoset composites with sub-1-minute dwell times; validation of impact performance and evaluation of the efficacy of the compression forming process. CRADA Final Report

Mallinda is developing polyimine malleable thermoset prepreg composite materials which have excellent mechanical properties (100 GPa tensile modulus, 2 GPA tensile strength, 2.4% elongation at break) and high operating temperatures (Tg>200°C). At scale, polyimine resins are commensurate in price with commodity epoxy resins. What distinguishes malleable thermoset prepreg from traditional thermoset prepreg materials, is that they are fully cured during Mallinda’s roll-to-roll production of prepreg laminate. This results in 5 key value-differentiating benefits. First, it simplifies manufacturing logistics by enabling ambient transportation and storage, and by significantly extending out-life and shelf-life almost indefinitely. Second, elimination of autoclave curing reduces the economic and energy costs to the customer. Third, scrap rates can be reduced as malleable thermoset prepreg materials are directly reusable. Fourth, the manufacturing consolidation step can be roughly 10x faster than traditional thermosets, because the resin is already cured. Parts can be made via compression forming by the application of heat and pressure to quickly vitrify and consolidate a multilayer part – easily leading to sub 3-minute cycle times (at lab scale we have demonstrated a 20 second dwell time, with room for further optimization). Finally, the closed-loop cradle-to-cradle solution-based recyclability of malleable thermoset composites can also contribute significantly to the future of sustainable lightweight materials. The focus of this project was the development, optimization and validation of malleable thermoset composite materials which exhibit manufacturing cycle times of 3-minute or less, high speed impact performance on par with incumbent technologies, and defect-free consolidation of 3 dimensional parts.

36 MATERIALS SCIENCE↗
NASA NTRS1980

Influence of Microstructure on the Fatigue Crack Growth of A516 in Hydrogen

Some day hydrogen may be used as a viable energy storage and transport medium within the United States. Hydrogen gas may be used to dilute and extend our present methane supply as a blend or may even be used in its pure elemental form as a primary fuel. Independent of the methods of production, storage, and distribution, the interaction of hydrogen with its containment material will play an integral role in the success of a hydrogen energy program. Presently, the selection of hydrogen containment materials can be made such that the material will remain reasonably free from environmental degradation; however, costly alloying additions are required. Unfortunately, high alloy steels are economically prohibitive when large-scale hydrogen energy storage, transmission, and conversion systems are desired. Therefore, in order to implement such hydrogen energy systems in the future, existing low-cost materials must be improved via mechanical, thermal, or thermo-mechanical processing methods or new low-cost materials which are compatible with hydrogen must be developed. Originally, low strength, low alloy steels at room temperature were thought to be immune to hydrogen gas embrittlement, since no sustained load crack growth is observed. However, results of Clark in HY8O and Nelson in SAE 1020 have shown that the fatigue crack growth rate can be greatly accelerated in the presence of hydrogen gas. In recent results reported by Louthan and Mucci, the smooth bar fatigue life of an A1068 pipeline steel was reduced up to a factor of ten when the tests were performed in a 13.8 MPa hydrogen environment. These results suggest that the selection of material for structures designed to operate in hydrogen under cyclic loads must include consideration of hydrogen/metal fatigue interaction. Although the hydrogen/metal fatigue interaction can be severe in low strength low alloy steels, the degree of degradation may be altered by the underlying ferrous microstructure. At present, no correlation between microstructure and degree of hydrogen susceptibility exists for low strength steels. However, in high strength steels, susceptibility to hydrogen embrittlement has been shown to be strongly sensitive to the metallurgical microstructure. In addition, compositional effects and grain size can 703 Some day hydrogen may be used as a viable energy storage and transport medium within the United States. Hydrogen gas may be used to dilute and extend our present methane supply as a blend or may even be used in its pure elemental form as a primary fuel. Independent of the methods of production, storage, and distribution, the interaction of hydrogen with its containment material will play an integral role in the success of a hydrogen energy program. Presently, the selection of hydrogen containment materials can be made such that the material will remain reasonably free from environmental degradation; however, costly alloying additions are required. Unfortunately, high alloy steels are economically prohibitive when large-scale hydrogen energy storage, transmission, and conversion systems are desired. Therefore, in order to implement such hydrogen energy systems in the future, existing low-cost materials must be improved via mechanical, thermal, or thermo-mechanical processing methods or new low-cost materials which are compatible with hydrogen must be developed.

Wachob, Harry F.↗
DOE OSTI2024

Coupling of the Electricity and Transportation Sectors - Part I: Sector Overviews

“Sector coupling” is a concept that addresses potential designs for the future power industry. Traditionally, the energy sectors, i.e., electricity supply, transport, and industry, have functioned largely independently from one another. Thus, the concept of sector coupling means that the electricity sector would become the central pillar of the energy system, supplying the other sectors – transportation and industry – with energy in various ways, a scheme described by the term “Power-to-X”. This report delves into the ambitious goal set by the United States to achieve net-zero emissions by 2050, with a pivotal milestone of making half of all passenger vehicles sold in America zero-emission by 2030. Central to achieving this objective is the electrification of the transportation sector, resulting in complex interactions between this sector and the electric industry. As electric vehicles become more prevalent, a shift in the dependency on electricity for vehicle charging emerges, alongside a reduction in the electricity demand associated with maintaining the fossil fuel supply chain. Both the electricity and fossil fuel have been identified as critical infrastructure sectors, the transportation sector's historical reliance on fossil fuels, and the mature supply chain supporting them are explored. This study emphasizes the existing sectoral coupling between the electric energy and transportation sectors, extending beyond vehicle charging to essential operations at refineries, pipeline facilities, storage, and fuel extraction. Recognizing the critical dependency of fossil fuel supply chain operations on the electric sector, this research underscores the need for a comprehensive view of the evolving dynamics between the transportation and electric sectors. This study has been split into two segments, Part-I provides an overview of the existing fossil fuel supply chain architecture and the electrical sector. We provide an in-depth analysis of the grid-transportation sectoral coupling, its interactions, the vulnerabilities and essential take-aways that would help us to address them and apply towards an electrified transportation sector in the future. We engaged with stakeholders to understand their expectations in enabling an electrified transportation. We accumulate the understandings from the current fossil fuel-based system in developing a risk matrix, by identifying all the threats and vulnerabilities and how would they apply to both the current and future transportation systems in Part-II of this report.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2020

HBU R and D Cask Transportation Options - 20436

The TN-32B HBU Demonstration Cask was successfully loaded in November 2017 with 32 high burnup (HBU) fuel assemblies at Dominion's North Anna Nuclear Plant and is collecting thermal data as intended. The first two phases of this very successful project sponsored by DOE and EPRI was designed to collect thermal data of HBU fuel to support extended storage. The next phase of the project requires the cask be licensed for transportation so the cask can be relocated to a site (to be determined by DOE) to facilitate long term monitoring and additional research on the fuel. This is a unique challenge given the constraints that exist for this particular bolted metal cask. The TN-32B cask is a 234,288 lb. (storage configuration) dry storage system that was selected for this project and it was modified to provide thermal data and the ability to collect gas samples after loading. Several major issues must be addressed to obtain the transport license including protecting the seven lid penetrations that were made to allow for the insertion of the thermocouples into the selected fuel assemblies after the cask was loaded. In order to protect the integrity of the research, the payload must remain as is. Therefore, the cask system must be placed into the transportation configuration without rewetting the fuel, replacing the lid and without removing the thermocouple instrumentation that is currently installed. A critical portion of the design and engineering that was performed in the first two phases of the project was to ensure that future transportation of the cask was integrated into the approach. Efforts were made to accommodate requirements such as containment boundary leakage testing, installation of impact limiter brackets and lid bolts suitable for transportation. Due to the science requirements of the project, the fuel selected was intended to yield the highest payload temperatures permitted to provide the greatest value for the researchers. This resulted in higher than usual external dose rates and surface temperatures that warranted performing as much preparatory work prior to cask loading to ensure ALARA and personnel safety to the greatest extent practical rather than working on a loaded cask. In addition to the work performed prior to cask load, there are remaining challenges such as protecting the modified lid, design and fabrication of impact limiters that accommodate the modified cask lid and preserving the thermocouple connections for continued monitoring. The project team is currently performing the necessary calculations, engineering, and design to support the licensing as required by the US NRC. Pre-application meetings have been held to inform the regulator of the planned technical approach to ensure compliance with the applicable regulations. The current plan is to obtain the transportation C of C by the end of 2021. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗