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MFC Transient Testing Capabilities and Fuel Fabrication capabilities are summarized in this slide deck. Slides are included for TREAT and TREAT capabilities, EFF, FASB, FMF, ZPPR, AFF, and the Fuels Fabrication Modernization Capability. First use of this slide deck is for presentation at Westinghouse Electric Corporation, but there will be other uses later.
Since the Manhattan Project, the U.S. Department of Energy (DOE) and its predecessor agencies have performed a wide mission involving nuclear materials and weapons development, production, testing, and modernization. While the initial mission components relied heavily upon reactor physics and chemistry, the environmental programs have increasingly relied upon development of more accurate and effective measurement methodologies to determine the radiological and chemical properties of the unique waste streams that were generated during the weapons production and testing mission. During the ongoing site cleanup program, the DOE laboratories and former production sites have developed a wide range of site-specific processes and procedures for conducting radiological characterization of dry active waste (DAW) that is managed as low-level radioactive waste. This has resulted in a Complex-wide variability among sites with respect to the methodologies and instrumentation used to characterize individual waste streams. The DAW-type waste was selected for development of a characterization standard since it is a simple, voluminous, and repetitive waste stream with a substantial knowledge base, established experience history at multiple site locations, and the potential for realization of programmatic benefits though standardization. Implementation of a consistent, standardized approach to DAW characterization will take advantage of multiple applications for a technology that is both mature and stable. The DAW characterization process is repetitive and lends itself to implementation at a wide range of operating locations - resulting in economies of scale, compliance with regulatory requirements, and consistent implementation of procedures which support design, operations, maintenance, and management of facilities, activities and environmental restoration programs. In response to a request from DOE, the Energy Facility Contractors Group (EFCOG) Waste Management Working Group (WMWG) is supporting the development of a new DOE Technical Standard for Radiological Characterization of DAW. The purpose of the proposed new standard is to provide a technical resource that integrates and enhances the utilization of current site-specific methodologies and processes. An EFCOG/WMWG technical team has prepared a Technical Basis Document (TBD) to support the development and utilization of the new standard. The TBD provides detailed reference information for use by site personnel when implementing the guidance to be provided in the standard. It also provides the rationale for and details concerning the recommended technical approaches for radiological characterization of representative DAW-type waste streams. The TBD is an essential resource and will be included as an appendix to the proposed standard. Formulation of the new standard will draw upon the information contained in the TBD in order to integrate and improve the utilization of current site-specific methodologies and processes. It will also include a comparison of current radiological characterization methodologies to determine which ones are most reliable and effective in producing relevant and valid data. In view of the long history and participation by several agencies in the regulation of nuclear programs, existing Voluntary Consensus Standards have also been identified in the TBD and will be referenced for relevancy and applicability to the new proposed standard. A Project Justification Statement (PJS) was also developed by the EFCOG/WMWG Team. Following internal Team review and approval by EFCOG, the draft PJS was submitted through Argonne National Laboratory (ANL) and the DOE Argonne Site Office Technical Standards Managers for review and approval by the DOE Argonne Site Office Manager. After resolution of comments and approval by the Manager, the PJS will be submitted to the DOE/HQ Technical Standards Program Manager (TSPM) for formal coordination. Following resolution of technical comments received during the PJS coordination process and receipt of a technical standards project number assigned by the TSPM, the Team will confirm the overall preparation schedule and develop an Internal Draft Technical Standard for coordination with the DOE Argonne Site Office. Following comment resolution, the Team will then prepare a Coordination Draft for submittal to the DOE Technical Standards RevCom review process and will respond to comments generated during this Complex-wide review. (authors)
This presentation touches on motivation and applications of Zirconium. Performance of evaluated libraries on double and quasi differential experiments is considered. Additionally, the performance of evaluated libraries in select ICSBEP benchmarks is considered. Provided is a detailed look in the performance of evaluated libraries in preliminary benchmark configurations for IER-516: Zirconium Test Assembly (ZTA). Conclusions and recommended experiments to satisfy Zirconium scattering nuclear data needs are presented.
This study presents scaled experiments using steam condensation with non-condensable gas (NCG)—helium, simulating hydrogen—as these experiments are pivotal for water-cooled reactor passive containment cooling system (PCCS) design and analysis. Research into PCCSs for small modular reactors (SMRs) is especially important in light of SMR system design; however, studies in the literature reflect limitations due to test geometry and operational condition variations, without considering SMR prototypic design. To address these challenges, a scaled test facility was developed to accurately replicate SMR PCCSs. This facility includes vertical down-flow condensing test sections with 1-, 2-, and 4-in.-diameter condensing tubes, accompanied by annular water cooling. Experiments were conducted using both superheated and saturated steam, with steam mass flow rates varying from 55 to 66 kg/hr., in the presence of helium as the NCG mass flow rate ranges from 1.8 to 22 kg/hr. Test data were collected on (a) the axial temperatures of the annular cooling water; (b) the outer wall temperature of the condensers; and (c) the mass flow rate, temperature, and pressure at the test section inlets and outlets. These primary test data were used in conjunction with a standard data reduction methodology to estimate essential thermal parameters such as heat fluxes, heat transfer coefficients, and condensation rates. The effects of NCGs on steam condensation within the geometry of the scaled test sections were then presented in regard to various testing conditions.
A long term research focus of the US Department of Energy (DOE) Advanced Fuels Campaign (AFC) is to investigate technologies that can allow for increased actinide utilization in nuclear fuel for fast neutron spectrum reactors. In order to reach these goals, many experimental campaigns have been tested in an approximated fast reactor neutron spectra at the Idaho National Laboratory Advanced Test Reactor (ATR) in order to study the performance of metallic, nitride, ceramic and advance fuel forms. In addition and complementary to ATR testing, systematic characterization of experiments irradiated in true fast reactors are performed in order to assess ATR in qualifying fast reactor fuel. For this purpose, complementary postirradiation examinations (PIE) are performed on several metallic fuel pins previously irradiated in the EBR II. Advanced PIE techniques developed since the initial PIE of EBR II samples in the 1980’s have been applied in new examinations of irradiated samples during past EBR-II operations. The EBR-II metallic fuel pins analyzed in this report are X501 G591, an early minor actinide-bearing metallic fuel alloy; X430-T653, a U-19Pu-10Zr metallic fuel alloy with a larger diameter compared to EBR-II driver fuel, and more prototypic of future commercial fast reactors; X521-G594, a unique experiment to test the feasibility to burn weapon grade Pu without reprocessing; X496 CL47&CL39 from a long-life metallic fuel experiment with the goal to reach very high burnup; and X486-J555 a U-10Zr Mk-IIIA qualification fuel test that will serve as first of a kind “modelling driven PIE.” These legacy pins were evaluated against the historical fuel performance of previously irradiated fuel from literature. Fuel performance was investigated through a suite of PIE techniques including neutron radiography, gamma spectrometry, dimensional inspection, fission gas release measurements, chemical burnup analysis, optical microscopy, and advance microscopy. So far, the PIE campaign revealed fuel performance and characteristics that are generally in line with literature or EBR-II experience. Some exceptions are present, and they are directly discussed in this report.
For any technology, the pace of development is determined in part by the time and expense required to explore alternatives to the current state of the art. For example, the capabilities of software are growing exponentially in large part because the costs of experimentation are so low (e.g., a programmer can often try out a new idea in a matter of minutes). For nuclear fuels and materials, the time and cost of iterating on alternatives is extremely high, which in turn helps to explain why nuclear reactor development can be excruciatingly slow. The pace of development, however, can be accelerated in two distinct ways: (1) improvements in the productivity of our irradiation testing and post-irradiation examinations infrastructure; (2) the use of a variety of alternatives to radiological work, including physically-based models, extensive literature reviews, increased reliance on out-of-reactor testing, and evaluation of the impact of fuel system developments on reactor plant performance. Viewed in this context, a fuel development program effort consists not only of test pin fabrication, irradiation testing, and post-irradiation examinations, but also on physically-based models, out-of-reactor testing, benchmarking of models against historic fuel performance data, and literature reviews. Of course, elements of these efforts are already in place, but increased emphasis on these activities are expected to speed the fuel development effort. Quite simply, the approach calls for obtaining as much knowledge and understanding as possible using whatever means necessary while minimizing the amount of radiological work that is required. In this manner, once an irradiation test program is designed and initiated, there is not only an improved likelihood that the fuel system will behave as expected but there is also a better chance, if the testing is successful, that the fuel design will result in substantial improvements to reactor plant economics and/or safety.
In the context of the development of Enhanced Accident Tolerant Fuels (EATF), AREVA NP, along with CEA and EDF, is developing, among others, a chromium (Cr) coating (5- 20 μm thick) on zirconium alloy cladding. The coating is deposited on the cladding surface by Physical Vapor Deposition (PVD) and forms a dense and adherent protective layer. It preserves the good mechanical behavior of the current cladding in normal operating conditions while significantly improving the oxidation behavior in accidental conditions. This behavior is being evaluated through out-of-pile corrosion and mechanical tests at the AREVA NP Technical and Research Centers. The corrosion behavior was evaluated through autoclave tests in both PWR water chemistry at 360 and in 415 Celsius degrees steam to increase the corrosion kinetics. In both cases very small weight gain was measured on the order of 1 mg/dm{sup 2} and remains stable throughout the length of the tests. The mechanical behavior of Cr-coated samples was evaluated through tensile tests on Cr-coated and uncoated M5 tube samples at both room temperature and 400 Celsius degrees. The curves demonstrate that the Cr-coated samples exhibit similar mechanical behavior as uncoated samples, and for all the tests performed, the mechanical properties (yield strength 0.2%, ultimate tensile strength, and total elongation) of the Cr-coated samples all fell within the range of values for uncoated samples. In order to obtain in-pile data as early as possible on this solution, an irradiation program (called IMAGO) in the Goesgen reactor will be performed in the years to come. This will be the first irradiation of EATF solutions in a commercial reactor.
In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. Furthermore, the designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability. Additionally, the TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.
This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.
Tri-structural Isotropic (TRISO) encapsulated fuel particles are used in high-temperature gas-cooled nuclear reactors. Fuel performance is limited by the propensity of metallic fission products to be released under high temperature testing to simulate hypothesized accident scenarios. The current particle design consists of a 350 μm or 425 μm UO{sub 2} or UCO fuel kernel surrounded by a 100μm porous carbon buffer layer, a 40μm inner pyrolytic carbon (IPyC) layer, a 35μm high-density silicon carbide (SiC) layer, and a 40μm outer pyrolytic carbon (OPyC) layer. Of interest is the fission product interaction with the SiC layer, which provides the most structural support and is the main barrier to the diffusion of fission products. Previous laboratory examination studies have indicated the release of Ag through seemingly intact particles and the localized corrosion of the SiC due to Pd. Although a number of studies have been conducted, the specifics of both the Ag and Pd reaction mechanisms, in addition to U and Pu, through SiC are currently not well understood. This work aims to understand the pathways and kinetics of these interactions under normal operation as well as accident conditions. In order to improve our understanding of this, we have the unique opportunity to examine irradiated TRISO particles to study post-irradiation effects on these materials. Experimental data is necessary to constrain the current models used to predict the behavior of fission products in the SiC layer of TRISO fuels. This work aims to understand the interaction of fission products with the SiC containment layer in TRISO particles. We have systematically examined three SiC shells from irradiated TRISO particles using XAFS, TEM, and EDS. Results indicate the formation of palladium silicides, and uranium and plutonium carbides. Additionally, silver is seen to be metallic. Preliminary microscopy results indicate fission products throughout the SiC layer of the TRISO particles, and microstructural changes have been observed between the as-irradiated and safety tested particles. These results have provided key information on the fission product transport behavior through irradiated SiC at varying temperatures. Knowledge of these reaction pathways will allow for better simulation of the long-term behavior of TRISO fuels. They may also suggest ways to modify the SiC layer to improve fuel performance and mitigate fission product release, which is critical for safety strategies required to commercialize these nuclear fuel technologies. (authors)
Environmental fatigue modeling is a complex problem due to multiple failure modes and their intermixing. The failure modes are function of various underlying causes in addition to the corrosive effect of reactor coolant environment. Some of the major causes are time-dependence of material associated with cyclic loading, load sequence effect associated with random/variable amplitude loading, effect of strain amplitude and rates, effect of varying temperature (along both temporal and spatial directions) and the effect of mean strain and stress. The nonlinear intermixing of failure modes associated with above mentioned causing parameters makes the environmental fatigue modeling is a challenging task. Because of this challenge, fatigue is traditionally being modeled based on experimental data. However, test based empirical approach often requires hundreds of fatigue tests to model the above-mentioned intermixing failure causes even for a single material system. The problem is further exaggerated for reactor component made from multi-material systems such as made from both carbon and stainless-steel base metals and their similar and dissimilar metal welds. With the difficulty of conducting hundreds of fatigue tests to capture the above-mentioned intermixing failure causes, fatigue modeling approaches often depends on empirical models based on limited available test data such as available through ASME code and NUREG 6909. However, these limited test-data-based models may not be enough to accurately predict the life of reactor components. Accurate prediction of life of reactor component would become a necessity, particularly when the license of the reactors to be extended for long-term-operation (LTO) that is for well beyond its original design life of 40 years. The requirement of extending the license of reactor under LTO requires hundreds of fatigue tests to be conducted to understand the mechanism associated with the above-mentioned interdependent failure causes. However, conducting large number of fatigue tests is not a feasibility due to the cost involved. To address this issues Argonne National Laboratory (ANL) with the sponsorship of DOE Light Water Reactor Sustainability (LWRS) program trying to develop a hybrid predictive modeling approach. This is based on limited experiment-data, Artificial-intelligence (AI) – Machine-Learning (ML) - Deep-Learning (DL) based techniques and Multiphysics-computational-mechanics based modeling tools. The hybrid approach not-only can improve the accuracy of the existing stress analysis and fatigue modeling approach but also can reduce the over-dependency on test-based approach. Towards this goal following are some of the major contributions based on ANL’s FY-20 environmental fatigue modeling activities: 1) A cyclic plasticity material model database for 82/182 dissimilar metal weld, which can be readily shared with US nuclear industry and regulatory agency on request. 2) A well validated analytical modeling methodology to perform cycle-by-cycle stress prediction under both constant amplitude fatigue loading and variable amplitude fatigue loading (with load-sequence effect). 3) An AI/ML/DL based methodology to predict unmeasurable cyclic strain based on other available sensor signals. This type of approach can be used for estimating strain in real reactor components from other sensor readings. 4) An AI/ML based approach to improve the US capability on environmental fatigue testing. This is by improving ANL’s existing environmental fatigue testing capacity to conduct ASME required strain-controlled tests (by controlling strain amplitudes and its rate), while not measuring the strain (due to the difficulty of placing an extensometer in a narrow autoclave in a PWR-water-test system). 5) A simulation and experiment based probabilistic modeling methodology for time-series fatigue state and life estimation of reactor metal such as dissimilar metal weld.
The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the analysis shows that the propagation of error method introduces essentially negligible error, especially when compared to the uncertainty associated with some of the estimates themselves. The results of these uncertainty analyses generally quantify and identify the sources of uncertainty in the overall cost estimation. The obvious generalization—that capital cost uncertainty is the main driver—can be shown to be an accurate generalization for the current state of reactor cost analysis. However, the detailed analysis on a component-by-component basis helps to demonstrate which components would benefit most from research and development to decrease the uncertainty, as well as which components would benefit from research and development to decrease the absolute cost.
Transient overpower testing for fuel safety research and development has resumed at Idaho National Laboratory. The THOR-C-2 commissioning test has been performed and post-transient examination has been performed revealing the intended creep rupture failure near the top of the fuel zone. This paves the way for future testing on previously irradiated fuels that will support fuel development and qualification.
The following report focuses on utilizing directed energy deposition (DED) for constructing the Transformational Challenge Reactor core. Multiple additive manufacturing processes are under investigation to evaluate the advantages and viability of each process for direct manufacturing of complex structures needed for future nuclear reactors. Technologies under investigation include binder jet, laser powder bed (LPB) fusion, electron beam powder bed fusion, and DED. Each process has advantages and disadvantages. For the core design, the current manufacturing approach consists of integrating LPB with DED. This hybrid manufacturing approach uses LPB for the core’s top and bottom manifold (where complexity is high) and DED for the manufacture of the cladding, where complexity is still high but geometric variability, size, and speed are better suited for DED. Research has focused on processing parameter development, evaluation of toolpath generation algorithms, core designs suitable for DED, manufacturing tolerances for integration of fuel elements, and seamless calibration and integration of LPB components into a DED build volume to create monolithic structures based on integrating these technologies. The following report highlights reactor core design as it relates to DED and experimental validation of test components necessary for finalizing the core design.
Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4 ) and silica (SiO 2 ) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.
Maintaining Continuity of Knowledge (CoK) of spent fuel throughout its life cycle is a key objective of the International Atomic Energy Agency (IAEA). This report describes a project in which two specific technologies, cameras with infrared (IR) imaging capabilities and imaging sonar, were tested to determine their effectiveness in augmenting optical surveillance cameras currently used for maintaining CoK for spent nuclear fuel (SNF) staged in spent fuel pools. Furthermore, the project identifies any modifications required to the technologies to facilitate deployment for international safeguards purposes. This report presents results for the imaging sonar tested at Sandia National Laboratories (SNL) in a mockup spent nuclear fuel pool and for the IR-capable cameras tested by Pacific Northwest National Laboratory (PNNL) in the Oregon State University Training, Research, Isotopes, General Atomics (TRIGA) research reactor.
Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on an organically cooled, heavy water moderated nuclear reactor in the world, Whiteshell Reactor 1 (WR-1). WR-1 was safely shut down in 1985, de-fueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed In Situ Decommissioning of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, pressure tubes) was necessary to verify inventory estimates and neutron activation models produced circa 1992. The Fuel Channels of WR-1 (pressure tubes) were expected to be the most activated components of the reactor core, and to account for about 60% of the reactor core inventory. The interior surface of the fuel channel is also the most readily accessible surface within the reactor core. As such the decision was made to collect a sample of one of each of 3 Fuel Channel material types (Stainless Steel, Zr-2.5Nb, Ozhenite) for comparison to modelled estimates of neutron activation of various core materials. The expected dose rates from the samples were between 5 and 100 mSv (500-10,000 mrem) on contact based on a 200 mg sample. Sampling the Fuel Channels required verification the Fuel Channel was drained of organic coolant, removal of residual organic coolant from the Fuel Channel wall, removal of material from the interior surface of the Fuel Channel at the centerline of flux (∼25' below the operator position), and extraction of that material from the core. Several options for existing tooling and technologies, including hydraulic scrapers, and horizontal drill technologies, were looked at for their applicability to WR-1. While suitable approaches to higher radiation field sample collection, they were overly complex, costly and time consuming for the configuration at WR-1. In an effort to reduce costs, and to simplify the sampling process, a specialized and simple tool was developed to mechanically scrape a small sample of the interior wall of the Fuel Channels, collect the sample into a vial, and remove the sample from the core for analysis. The tool was designed, constructed, tested and implemented in the field entirely by Canadian Nuclear Laboratories Staff. The tool required no powered components, or electronic guidance or measurements; and was designed and constructed for a fraction of the cost of other commercially available technologies for sample collection within reactor cores. The samples were successfully collected without any injuries, contamination events or equipment failures. The tool remains operational in the event additional sampling is required. This paper provides additional details of the design, construction, testing and use of the tool and the successes and lessons learned during the various stages of design and implementation. (authors)
The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of accident tolerant fuel (ATF) in United States commercial nuclear power reactors. Pacific Northwest National Laboratory is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs. This report focuses specifically on the iron-chromium-aluminum (FeCrAl) alloys being investigated to replace zirconium-based alloys for fuel cladding and provides current state-of-the-industry information on material properties and fuel performance considerations under reactor operating conditions and design basis accident conditions. Currently, Global Nuclear Fuels (GNF) is the only U.S.-based fuel vendor considering FeCrAl cladding as a near-term ATF design. GNF has tested several different FeCrAl alloys including Kanthal APMT, C26M, and MA956; fueled C26M rods and unfueled Kanthal APMT, C26M, and MA956 rods have been inserted in lead test assemblies at Plants Hatch and Clinton. This report will focus on those three alloys specifically, with broader information given regarding in-reactor and ex-reactor testing of FeCrAl alloys. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of FeCrAl cladding concepts, including fuel performance characteristics that may not be addressed within existing regulatory documents.