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Iron-Chromium-Aluminum Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement
To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing iron-chromium-aluminum (FeCrAl) alloys in place of zirconium-based alloys in major core structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.
Cr-coated Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement
To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing the chromium-coating on major zircaloy structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.
United States Department of Energy severe accident research following the Fukushima Daiichi accidents
The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCOR results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U. S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. Finally, this paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.
Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program
The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes chromium (Cr) coated zirconium alloy cladding with uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. The lead test assembly (LTA) program will use both SiC/SiC composites and Cr coated zirconium alloy claddings with the high density/high thermal conductivity U 3 Si 2 pellet which will begin in 2022. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in autoclaves and in the Massachusetts Institute of Technology (MIT) reactor and the U 3 Si 2 pellets have been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. Fuel rod and assembly design in preparation for the LTR and LTA programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Cr coated zirconium alloy claddings showed low corrosion rates in both autoclave and in-reactor tests and will be fabricated using commercial processes for the first LTRs for insertion in 2019. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year which meets corrosion requirements may be achieved under normal operating conditions. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that solid fission products can be contained within SiC cladding for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher temperature capability of SiC (~2000°C). These two hours can be used to implement additional response options instituted through the FLEX program by reactor operators. While the Cr coated Zr alloy option offers more modest ATF gains (~200°C higher than bare Zr) in similar situations, the coatings do delay the runaway oxidation encountered by uncoated zirconium cladding. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely as long as the water flow continues. SiC exhibits the highest accident tolerance of any cladding and significant advances in understanding the design, manufacture and accident behavior of SiC cladding have been made. Department of Energy support is needed to continue the on-going experimental work addressing the interaction between U 3 Si 2 and SiC at temperatures >1200°C, the effect of SiC manufacturing methods and manufacturing variability on physical properties, and the ability to predict the mechanical properties of SiC composites based on the composite design. Additionally, experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 to 50 MWd/kgU are underway. These experiments will provide the data needed to license fuel rod design codes. Research is continuing on production methods for making U 3 Si 2 from UF 6 without going through the U metal step. Oxidation tests of fuel pellets and powder in steam and synthetic air indicate that U 3 Si 2 has a lower oxidation reaction initiation temperature than UO 2 . Options to reduce the U 3 Si 2 oxidation rate are being explored. A hybrid fuel design using pellets with the weight ratio of 30%/70% U 3 Si 2 /UN is currently undergoing testing in the ATR. This approach is being explored to further increase the positive economics of the U 3 Si 2 fuel. In addition to the work supported by the USDOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings are organized by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. Since SiC, coated cladding and high density fuel options are not currently used in the nuclear industry, support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/ UO 2 fuel. There are no showstoppers for accident tolerant fuel technologies when considering supply chain, fabrication, enrichment suppliers, production, and licensing. Due to the cost savings from higher density fuel and the accident tolerance of coated claddings, Exelon has identified potential commercial plants to host the LTRs and LTAs. Planning is underway to implement this schedule.
Dynamic PRA Methods to Evaluate the Impact on Accident Progression of Accident Tolerant Fuels
Accident tolerant fuels (ATFs) are new nuclear fuels developed in response to the accident at the Fukushima power station in March 2011. The goal of ATFs is to withstand accident scenarios through better performance compared to currently employed fuels (e.g., small-scale hydrogen generation). This paper targets a method for evaluating and comparing ATF performance from a probabilistic risk assessment (PRA) perspective by employing a newly developed combination of event trees and dynamic PRA methods. Compared to classical PRA methods based on event trees and fault trees, dynamic PRA can evaluate with higher resolution the safety impacts of physics dynamics and the timing/sequencing of events on the accident progression without the need to introduce overly conservative modeling assumptions and success criteria. In this paper, we analyze the impact on the accident progression of three different cladding configurations for two initiating events [a large break loss-of-coolant accident (LB-LOCA) and a station blackout (SBO)] by employing dynamic PRA methods. The goal is to compare the safety performance of ATFs (FeCrAl and Cr-coated cladding) and the currently employed Zr-based clad fuel. We employ two different strategies. The first focuses on the identification of success criteria discrepancies between the accident sequences generated by the classical PRA model and the set of simulation runs generated by dynamic PRA using ATF. The second one, on the other hand, directly uses dynamic PRA to evaluate the impact of timing of events (e.g., recovery actions) on accident progression. In conclusion, by applying these methods to the LB-LOCA and SBO initiating events, we show how dynamic PRA methods can provide analysts with detailed and quantitative information on the safety impact of ATFs.
Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program (Rev. 2)
The beyond design basis accident at Fukushima Daiichi in March 2011 and the Three Mile Island accident in 1979 imply that severe nuclear accidents may occur at somewhat higher frequencies than previously predicted, and that the financial liabilities of such accidents can cripple a utility. While current fuel is more than adequate for design basis accidents, the potential for harm from beyond design basis accidents (BDBA) has been a driver for Westinghouse development efforts since 2003. After Fukushima, other fuel vendors also began pursuing new fuel materials that can provide significant increases in the time, enabling reactor operators to respond to unforeseen events before significant releases of fuel materials and fission products occur. In order to be commercially successful, accident tolerant fuel (ATF) products must also provide significant fuel cost savings, operating cost improvements as well as safety improvements. This report documents the engineering, design and testing that occurred from October 1, 2016, through September 30, 2018, by the Westinghouse team. In addition, it includes: 1. The results of modeling and simulation work completed to date that supports technical feasibility of the ATF concept in meeting the goals of the enhanced ATF program including a core level analysis predicting how an existing LWR will perform when loaded with ATF. 2. Regulatory Analysis showing path to LTA/LTR acceptance. 3. Business Analysis showing commercial viability of the ATF concept.
Transient performance of accident tolerant fuel in design basis accidents
A set of transient analysis has been performed to assess the performance of candidate accident tolerant fuel (ATF) against reference UO{sub 2} fuel with Zircaloy cladding under design basis accident (DBA) conditions. This study examined two ATF options, U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding. The accident analyses were conducted using the system code TRACE for a three-loop PWR plant. PARCS standalone full-core calculations were done for each fuel/cladding combination to provide inputs to the TRACE point-kinetics model. The design basis accidents analyzed were a reactivity insertion accident (RIA) and a large break loss-of-coolant accident (LBLOCA). The RIA was simulated by the insertion of a reactivity ramp of $1 (one dollar) in 0.1 s at hot full power (HFP) condition. The LBLOCA was a simulation of a double-ended pipe break in one of the cold legs of the primary loops. Differences in thermal responses of the fuel and the cladding are attributed to their thermal properties. The U{sub 3}Si{sub 2} fuel with its lower volumetric heat capacity and higher thermal conductivity (i.e. higher thermal diffusivity) generally results in a lower fuel temperature than the UO{sub 2} fuel and a correspondingly faster thermal response. Additional calculations for the U{sub 3}Si{sub 2} fuel shows that in a RIA the cladding material, SiC versus Zr, has little impact on the maximum fuel centerline temperature and the PCT (peak cladding temperature). An increase in the hot rod power factor is shown to raise the centerline fuel temperature in a RIA but has little impact on the PCT. The analysis results show that for a relatively fast decay heat driven transient, such as the LBLOCA, the thermal-hydraulics and fuel/cladding thermal properties dictate the overall response more so than the reactivity feedback. The results show that the PCTs in a LBLOCA are similar for all fuel/cladding combinations considered in this study. The LBLOCA analysis also notes the impact of the asymmetry in the system response. The fuel rods in the sector of the core connected to the hot leg of the broken loop (Sector 1 in the PWR model) is shown to have a much lower PCT than fuel rods in the other five sectors of the core. Results of this study indicate that for the two DBAs considered (RIA and LBLOCA) the transient behavior and the resulting peak cladding temperature of the two ATF options (U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding) are similar to the standard light-water reactor fuel (UO{sub 2} fuel with Zircaloy cladding). (authors)
Pellet cladding mechanical interaction as a potential failure mechanism during a control rod drop accident in a boiling water reactor
Boiling water reactors (BWRs) represent approximately one-third of the operating fleet in the United States, contributing significantly towards the global effort in reducing carbon emissions. Even though most of the operating fleet has been in operation for quite some time, continued advancements in new nuclear fuel (such as accident tolerant fuel) or operating regimes (such as power up-rates and higher burnup operation) necessitates similar advancements in modeling and simulation capabilities. Bison, a component of the Virtual Environment for Reactor Applications (VERA), is a high-fidelity fuel performance code able to explore the fuel performance of a wide variety of fuel types in one-, two-, and three-dimensions. Until recently, the code had not been used for analyses of BWRs. Modeling capabilities have been added for Gd-bearing UO{sub 2} and pure zirconium liners. New models have been added based upon the U.S. Nuclear Regulatory Commission (NRC) guide-lines for hydrogen pickup in Zircaloy-2 claddings and pellet-clad mechanical interaction (PCMI) failure during a reactivity insertion accident (RIA), known as a rod drop accident (CRDA) in BWRs. Implementing and/or improving Bison modeling capabilities extended its analytical reach to areas beyond its original intended purpose. This paper demonstrates one of these capabilities as a proof of concept. Recently developed models enable Bison to provide an alternate approach for cladding integrity determination in CRDA evaluations, which currently use bounding conservative estimates. As part of this demonstration, NRC guidance on hydrogen-pickup and PCMI failure models were utilized in this research. Even though more research is needed in establishing right inputs and process in this area, this paper demonstrates Bison's ability to determine cladding integrity in a CRDA evaluation. This first of a kind demonstration is a proof of concept in this area, which could potentially be extended to a number of other areas where a more accurate cladding integrity determination would be needed. (authors)
An Application of CICCT Accident Categories to Aviation Accidents in 1988-2004
Interventions or technologies developed to improve aviation safety often focus on specific causes or accident categories. Evaluation of the potential effectiveness of those interventions is dependent upon mapping the historical aviation accidents into those same accident categories. To that end, the United States civil aviation accidents occurring between 1988 and 2004 (n=26,117) were assigned accident categories based upon the taxonomy developed by the CAST/ICAO Common Taxonomy Team (CICTT). Results are presented separately for four main categories of flight rules: Part 121 (large commercial air carriers), Scheduled Part 135 (commuter airlines), Non-Scheduled Part 135 (on-demand air taxi) and Part 91 (general aviation). Injuries and aircraft damage are summarized by year and by accident category.
System code evaluation of near-term accident tolerant claddings during pressurized water reactor station blackout accidents
Following the Fukushima Daiichi nuclear accident in 2011, researches on Accident-tolerant fuels (ATFs) are currently of high interest in not only the nuclear industry but also governmental and international organizations. In this work, a quantitative evaluation of the performance of monolithic FeCrAl cladding and Cr-coated Zircaloy cladding has been performed for Pressurized Water Reactor (PWR) Station Blackout (SBO) accidents. A generic PWR model has been built in system thermal-hydraulics code TRACE based on the Surry Nuclear Power Station with counter-current natural circulation modelling capability for hotleg and steam generator U-tube components during the accidents. The base model results are then compared to MELCOR and RELAP simulations to verify the system component implementation in TRACE. Two PWR SBO scenarios were investigated, including: short-term SBO and long-term SBO with early reactor coolant pump (RCP) seal failure. These scenarios are defined to be very similar to the accidents studied in the State-of-the-Art Reactor Consequence Analysis (SOARCA) project. TRACE code is modified to reflect the oxidation kinetics of FeCrAl and Cr-coating. Larson-Miller creep rupture model is also implemented in TARCE using its built-in control systems to simulate the creep rupture of hotlegs. Additionally, the comparison between the TRACE models with and without the counter-current flow modeling resulted in significant difference when comparing ATF cladding to Zircaloy for short term SBO, while it marginal impacted the performance during long term SBO with RCP seal failure. For short term SBO, both ATF cladding underwent hot leg creep rupture ~20 min after Zircaloy cladding. While Zircaloy and Cr-coated cladding had generated significant amount of hydrogen gas (>10 kg) before the creep rupture event, FeCrAl cladding had only generated <0.5 kg of hydrogen gas. For long term SBO with RCP seal failure, significant hydrogen generation and fuel melting was predicted before hot leg creep rupture for the ATF cladding while providing only 10–20 min additional coping time compared to Zircaloy.
Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials
A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO 2 – Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This article describes modifications made to a version of the MELCOR code by the Idaho National Laboratory (INL) that allows its application to alternate fuel-cladding systems. Preliminary analysis results are presented in this paper for candidate silicon carbide and FeCrAl cladding concepts.
Development of Accident Tolerant Fuel Options for Near Term Applications: Fuel Performance Modeling under Transient / Severe Accidents by MELCOR: PART I: Benchmark for SURRY: Short Term Station Black Out (STSBO)
Currently, existing LWR fuel technology is near its inherent limit of performance without improvement in fuel materials. As a result of the great earthquake of east Japan, along with the catastrophic tsunami, the Fukushima Daiichi boiling water plant was severely damaged. The light water reactors (LWRs) fuel gave rise to a serious discussion in the United States. In order to provide the development of Accident Tolerant Fuel (ATF) options for present day applications, MIT is leading a DOE project to search for enhanced accident tolerant fuels. Group members include MIT, UW-Madison, TAMU, AREVA, and Westinghouse. As part of the work, fuel performance modeling under transient / severe accidents: MELCOR is being used to analyze different kinds of materials under several limited accident conditions. While various research groups are obtaining the parameters of the experimental ATF materials, benchmarks for SURRY are being done by use of MELCOR in the University of Wisconsin-Madison. We chose version 1.8.6 for the edition of materials, and INL provided the uncertainty analysis between version 1.8.6 and version 2.1.0. Comparison will be done between potential ATF materials, which include FeCrAl, SiC, Cr, ZrSi, TiAlC, TiSiC and so on. Furthermore, zircaloy is used as the verification. Among all the parameters which influence the materials' performance, the temperature history and hydrogen generation will be the key factors in the future. (authors)
Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident - 2016-0090
Along with our colleagues at the Japan Atomic Energy Research Institute (JAERI) and Japan Atomic Energy Agency (JAEA), we have performed extensive research programs for more than 2 decades and have developed a better understanding of fuel behavior under accident conditions. Data and findings from these research programs have provided the technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcomes from the research programs performed at JAERI and JAEA facilities and identifies further research needs. The programs aim to generate data and models to (1) evaluate the adequacy of present safety criteria and safety margins; (2) create a database for the future regulation of higher-burnup UO{sub 2} and mixed-oxide (MOX) fuels, new cladding, and pellets; (3) provide adequately mechanistic computer codes for regulatory application; (4) promote a better understanding of phenomena that appear in high-burnup regions and MOX fuels, such as rim effect and an effect of plutonium agglomerates in MOX fuels, and to evaluate those effects on fuel behavior in accident conditions; and (5) to assess fuel behaviors with higher fuel duty, such as plant power uprates, longer operating cycles, and water chemistry changes. The programs comprise reactivity-initiated accident studies, including pulse-irradiation experiments in the Nuclear Safety Research Reactor, cladding mechanical tests, and the development and verification of RANNS; loss-of-coolant accident tests, including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6 that simulates high-burnup fuel behavior under normal operating and abnormal transient conditions; and studies on phenomena specific to high-burnup regions, including thermal property measurements, experimental and analytical simulation of rim structure formation, cladding mechanical tests on effects of hydrides, etc. (authors)
New intrinsically smart severe accident instrumentation saves costs and enhances severe accident management - 302
The Severe Accidents (SA) at Fukushima in 2011, identified the need for enhanced I and C to follow the progress of SA. Since SA environments are significantly Beyond DBA conditions, some new advanced Inherently Smart sensors had to be developed and tested. Nuclear plant designers and operators have helped focus the I and C enhancement based upon lessons learned from the Fukushima accidents. The result has been the addition of I and C for direct measurement of SA conditions hydrogen concentration, oxygen concentration, temperature, and pressure that provide unambiguous information for management of Severe Accident scenarios. The benefits are immediate from three important aspects. First, they deliver on the nuclear promise for enhanced safety at reduced costs. These Inherently Smart sensors have no maintenance and replace equipment that has high maintenance costs. Secondly, these sensors provide multiple benefits, such as condition monitoring information all the time and direct measurement of SA conditions if they are needed. Thirdly, and most importantly, simplify operator actions and training for SA. SA scenarios and SA Management can now rely on direct measurements to know such things as, the extent of fuel damage, integrity of the reactor pressure vessel, when corium is on the floor of containment, and conditions upon which to activate the hardened vent. (authors)
Neutron cross section sensitivity and uncertainty analysis of candidate accident tolerant fuel concepts
Not Available
Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario
Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.
Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions
The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of the concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.