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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

AREVA enhanced accident tolerant fuel program - Current results and future plans

The purpose of this paper is to provide a brief overview of the enhanced accident tolerant fuel (EATF) concepts investigated by the AREVA team in Phase 1 to support the U.S Department of Energy Enhanced Accident Tolerant Fuel program and concerning light water reactors. The main goal of the 'fuel pellets with additive' concept is to improve the economics, performance and operational flexibility of fuel in light water reactors while featuring improved safety performance under normal and off-normal conditions. Within this concept 2 nuclear fuels were explored: Cr{sub 2}O{sub 3} doped UO{sub 2} pellet and composite UO{sub 2} pellets. The second concept is the coating of zirconium cladding whose main goal is to reduce cladding oxidation and hydrogen generation during accident conditions. The coating materials were chosen to be compatible with the specific deposition process, to provide robustness in the nuclear environment and have a relatively small neutron cross section. Three coatings were explored: chromium coated claddings, electrophoretic deposition of coating, and MAX phase coatings. MAX phases are a family of the layered compounds with a general formula M{sub n+1}AX{sub n} with n = 1, 2, or 3, where M is a transition metal, A is an A-group element, and X is either C or N. Another improvement to the fuel cladding that was considered is to replace the zirconium cladding altogether with an alternative cladding material which can substantially improve reactor response during postulated accident conditions. Two concepts were investigated: molybdenum claddings and SiC/SiC composite cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Bulk and surface controlled diffusion of fission gas atoms

Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. . One of the most commonly used models in fuel performance codes was published by Massih and Forsberg. This model is similar to the early Booth model in that it applies an equivalent sphere to separate bulk UO 2 from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D 3 ) and intermediate (D 2 ) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's and Matzke's data. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO 2±x , which compare favorably to available experiments. This is an extension of previous work. In particular, it applies improved chemistry models for the UO 2±x nonstoichiometry and its impact on the fission gas activation energies. The derivation of these models follows the approach that used in our recent study of uranium vacancy diffusion in UO 2 . Also, based on the calculated DFT data we analyze vacancy enhanced diffusion mechanisms in the intermediate temperature regime. In addition to vacancy enhanced diffusion we investigate species transport on the (111) UO 2 surface. This is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO 2 under irradiation, for which surface diffusion could be the rate-limiting transport step. Diffusion of such bubbles constitutes an alternative mechanism for mass transport in these materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Performance of Waterlogged LWR fuel during RIA

Fuel pins with small defects could allow for the gradual ingress of water into the cladding during normal operation. If these 'waterlogged' fuel pins are subjected to a rapid power change, the internal pressurization could greatly exceed that of a normal fuel pin due to rapid conversion of the water to steam. This could induce either a premature or a more energetic failure during a design basis RIA than in non-defected fuel pins. Experimental studies on this type of fuel are very limited and do not appear to have been fully reported in the open literature. Results from a variety of tests conducted at the Transient Reactor Test Facility (TREAT), Special Power Excursion Reactor Test (SPERT) facility, Nuclear Safety Research Reactor (NSRR), and Power Burst Facility (PBF) are reviewed in this paper. These tests were intended to evaluate both the physical consequences of failure and the threshold and causes for waterlogged fuel failure during RIA. The available data suggests that although waterlogged fuel pins may burst at a lower energy deposition threshold, they do not pose a significant new hazard during design basis accidents. However, data is limited to tests conducted on fresh fuel samples only.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Advanced Characterization of Fuel Pellet and Cladding from the HB Robinson NPP

In order to predict the behavior and response of materials to extreme environments of a nuclear reactor, comprehensive models have been developed over the past several decades. These models are typically validated against carefully designed single or double variable experiments which may be systematically altered in order build up a model of behavior under the conditions being examined. In reality, this does not necessarily compare well with the real application situations whereby the behavior of the material under irradiation is determined by multiple variables changing simultaneously e.g. temperature, irradiation flux and fluence, and pressure. This is particularly so when considering the highly dynamic environment of a fuel assembly within a nuclear power plant (NPP) core. This work aims to provide near-atomic resolution microstructural and chemical data of a fuel/clad specimen that has been used in service within a commercial NPP. Atom probe tomography and S/TEM characterization of the cladding showed the presence of small scale clusters containing Fe and Cr, along with segregation of these minor solutes to dislocations. SEM examination of the fuel pellet showed relatively large fission gas bubbles that had burst due to the metallographic polishing process to prepare the sample for examination. SEM-EDX mapping revealed the presence of Xr bubbles located beneath the surface that still retained their Xe concentration. Metallic precipitates were also observed. S/TEM-EDX of the fuel material also revealed the presence of small metallic precipitates. In addition, small Xe bubbles were also found to be present in the material, lying along grain boundaries only. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Problem Specification for FY12 Modeling of UNF During Extended Storage

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the Advanced Modeling and Simulation Office (AMSO) of the US Department of Energy, Office of Nuclear Energy (DOE/NE) has invested in the initial extension and application of advanced nuclear simulation tools to address relevant needs in evaluating the performance of used nuclear fuel (UNF) during extended periods of dry storage. There are many significant challenges associated with the prediction of the behavior of used fuel during extended periods of dry storage and subsequent transportation. The initial activities are focused on integrating with the Used Fuel Disposition (UFD) Campaign of the DOE/NE and a demonstration that the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) for modeling the mechanical state of the cladding after decades of storage. This initial focus will model the long-term storage of the UNF and account for the effect, and generation, of radially and circumferentially oriented hydride precipitates within the cladding and predict the end of storage (EOS) mechanical state (stress, strain) of the cladding. Predicting the EOS state of the cladding is significant because it (1) provides an estimate of the margin to failure of the cladding during nominal storage operation and it (2) establishes the initial state of the fuel for post-storage transportation. Because there are significant uncertainties associated with the storage conditions, hydride precipitate formation, and the beginning of storage (BOS) condition of the UNF, this will also allow for the development of a rigorous capability to evaluate the relative sensitivities of the uncertainties and can help to guide the experimental and analysis efforts of the UFD Campaign. This document is focused on specifying the problem that will be solved with AMPFuel. An associated report, documents the specifics of the constitutive model that will be developed and implemented in AMPFuel to account for the presence and predict the generation of the hydride precipitates. This report satisfies the deliverable for the DOE Office of Nuclear Energy, Advanced Modeling and Simulation Office, milestone M3MS-12OR0605083, 'Definition of Problem Specification,' which defines the problem to be solved that will satisfy milestone M2MS-12OR0605081, 'Demonstration of the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code for modeling UFD.' This document should provide sufficient detail to model a high burnup pressurized water reactor (PWR) fuel rod to provide an estimate of the end of storage (EOS) mechanical state of the cladding. The fuel rod and irradiation history are based on seven cycles of irradiation in the CP and L H.B. Robinson nuclear reactor, which achieved a discharge burnup of 66.682 MWd/kgU. The fuel has been experimentally examined for storage conditions by Argonne National Laboratory for the NRC. In addition, we have compiled a list of key factors that have been shown to strongly influence the EOS state of the fuel and have identified baseline values and ranges of uncertainties that will be considered. The simulations that will be performed have been described in detail and include the modeling assumptions and boundary conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

High-performace cladding-pumped erbium-doped fibre laser and amplifier

We report cladding-pumped erbium-doped fibre laser and amplifier configurations. Through fibre design optimisation, we have achieved a record-high laser slope efficiency, 40 % with respect to absorbed pump power ({lambda} = 976 nm), and an output power of 7.5 W. The erbium-doped fibre amplifier efficiency reaches 32 %.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 and this facility continues to support those efforts in the ATF community. Compared to the current UO 2 /Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Comparison of Subdivided Versus Full-Length Fuel Pin Burnup in the ATF-2 Safety Analysis

The Accident Tolerant Fuels (ATF) campaign, under the Fuel Cycle Research and Development (FCRD) Advanced Fuels program, was developed after the Fukushima Daiichi Nuclear Power Plant disaster in 2011. Under ATF, fuel and cladding concepts are to be developed and tested that may more readily withstand an accident similar to the one that occurred at Fukushima Daiichi. ATF is a combination of industry, universities, and national laboratories in an effort to develop advanced accident tolerant cladding and fuel concepts. The end goal of the ATF program is to insert a lead test rod (LTR) or lead test assembly (LTA) into an operating commercial power light water reactor by 2022. In order to meet this 2022 insertion goal, ATF is split into three separate phases: ATF-1, ATF-2, and ATF-3. The first phase of the ATF program, ATF-1, is the concept feasibility study and early irradiation testing. Feasibility, fabrication, and out-of-pile testing are happening at universities and laboratories all over the United States. Irradiation testing of these concepts is happening at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Early irradiation testing involves testing clad and/or fuel in drop-in capsules in the ATR. Currently, these capsules are in the relatively low-flux positions in the ATR, but some newer concepts and concepts that require higher burnup will be moved to the higher-flux positions later on during irradiation. Currently, Oak Ridge National Laboratory (ORNL), Westinghouse Electric Company, Areva, and General Electric (GE) have inserted capsules for ATF-1, although more concepts are planned. The second phase of ATF, ATF-2, is the focus of this paper. ATF-2 is the loop testing of fuel pins with new cladding and fuel concepts to simulate pressurized water reactor (PWR) and boiling water reactor (BWR) environments. ATF-2 will concurrently be happening at the ATR and at the Halden Reactor in Norway. The ATF-2 irradiation at the ATR will only simulate a PWR environment. ATF-2 will be inserted into the ATR pressurized water loop, Loop 2A, in the Center Flux Trap (CFT), where chemistry, temperature, and pressure can all be controlled separately from the primary coolant system and other loops in ATR. Loop 2A will be borated, pressurized to 2200 psi nominal, and have an average water temperature of 300 deg. C. A borated loop with fueled rodlets is extremely rare for Nuclear Energy experiments in the ATR, so ATF-2 will demonstrate new and unique capabilities for INL. ATF-3 is the testing of fresh and irradiated ATF concepts at the Transient Reactor Test (TREAT) Facility. The TREAT Reactor is planned for restart, re-establishing U.S. transient testing capabilities after nearly 20 years of non-operations. TREAT will test ATF concepts under transient conditions, such as a reactivity insertion accident. In addition to fresh fuel pins, TREAT will use from ATF2 and do transient testing on those irradiated pins. The information TREAT will provide to the vendors to support fuel licensing will be valuable, showing that these concepts do enhance the accident tolerance of the fuel in commercial reactors. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Corium lavas: structure and properties of molten UO 2 -ZrO 2 under meltdown conditions

In the exceedingly rare event of nuclear reactor core meltdown, uranium dioxide fuel reacts with Zircaloy cladding to produce eutectic melts which can subsequently be oxidized by coolant/moderator water. Oxidized corium liquids in the xUO 2 ·(100 – x)ZrO 2 system were produced via laser melting of UO 2 -ZrO 2 mixtures to temperatures in excess of 3000 K. Contamination was avoided by floating the droplets on a gas stream within an aerodynamic levitator and in-situ high-energy x-ray diffraction experiments allowed structural details to be elucidated. Molecular dynamics simulations well reproduced diffraction and density data, and show less compositional variation in thermal expansion and viscosity than suggested by existing measurements. As such, corium liquids maintain their highly penetrating nature irrespective of the amount of oxidized cladding dissolved in the molten fuel. Metal-oxygen coordination numbers vary with both composition and temperature. The former is due to mismatch in native values, n UO (x = 100) ≈ 7 and n ZrO (x = 0) ≈ 6, and the requirement for oxygen site stabilization. Furthermore, the latter provides a thermal expansion mechanism.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Severe Accident Test Station (SATS) Activity Report

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO 2 /Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000°C compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

Uranium silicide (U 3 Si 2 ) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U 3 Si 2 , including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U 3 Si 2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U 3 Si 2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U 3 Si 2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U 3 Si 2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U 3 Si 2 fuel as an accident-tolerant alternative for uranium dioxide.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution in the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A MULTIDIMENSIONAL AND MULTIPHYSICS APPROACH TO NUCLEAR FUEL BEHAVIOR SIMULATION

Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non-axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our the multidimensional, multiphysics approach to analyze a missing pellet surface problem. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Potential mechanism for lanthanide transport in metallic fuels

An experimental and modeling effort to determine the solubility and diffusivity of certain lanthanides in liquid metals has been investigated to assess the feasibility of liquid-like lanthanide transport mechanism in metallic fuels. We developed a conceptual pore model to better describe the lanthanide transport behavior through the fuel. Our results suggests that isolated pores will act most likely as lanthanide sink and the transport mechanism of lanthanides from fuel to the cladding has to occur through interconnected pores and cracks filled by sodium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A survey of alternative once-through fast reactor core designs

Reprocessing of Light Water Reactor (LWR) spent fuel to recover plutonium or transuranics for use in Sodium cooled Fast Reactors (SFRs) is a distant prospect in the U.S.A. This has motivated our evaluation of potentially cost-effective operation of uranium startup fast reactors (USFRs) in a once-through mode. This review goes beyond findings reported earlier based on a UC fueled MgO reflected SFR to describe a broader parametric study of options. Cores were evaluated for a variety of fuel/coolant/reflector combinations: UC/UZr/UO{sub 2}/UN;Na/Pb; MgO/SS/Zr. The challenge is achieving high burnup while minimizing enrichment and respecting both cladding fluence/dpa and reactivity lifetime limits. These parametric studies show that while UC fuel is still the leading contender, UO{sub 2} fuel and ZrH 1.7 moderated metallic fuel are also attractive if UC proves to be otherwise inadequate. Overall, these findings support the conclusion that a competitive fuel cycle cost and uranium utilization compared to LWRs is possible for SFRs operated on a once-through uranium fueled fuel cycle. In addition, eventual transition to TRU recycle mode is studied, as is a small test reactor to demonstrate key features. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Advancements in FeCrAl Alloys for Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors

Advanced FeCrAl alloys are being considered as an enhanced accident tolerant fuel (ATF) cladding for light water reactor (LWR) applications as they can exhibit slower reaction kinetics with high temperature steam and have a reduced production of hydrogen during accident scenarios compared both to historic 304L cladding and Zircaloy. Both of these attributes are key towards producing a fuel-form with enhanced accident tolerance over the current Zr-based-UO{sub 2} fuel form in deployment. A multi-year, multi-laboratory effort has been established to rapidly advance FeCrAl alloys as an ATF concept. This effort has focused on optimizing the microstructure and composition of these alloys in an attempt to provide performance on-par with or better than the current Zr-based cladding in both normal operating conditions and under accident scenarios. To date, a significant database supporting key performance factors in both scenarios has been developed. Based on this database, a compositional and microstructural window has been identified providing the insight for an optimized, advanced nuclear-grade FeCrAl alloy for LWRs. This work will provide a comprehensive overview of the systematic effort to advance FeCrAl alloys as an enhanced ATF cladding in commercial LWRs. Efforts undertaken by this multi-year, multi-laboratory program have shown the viability of using FeCrAl alloys as an ATF cladding in LWRs. Most importantly, it has been shown the oxidation resistance, decreased hydrogen production, and reduced heat generation of FeCrAl alloys strongly contribute to its ATF capability. Furthermore, it has been shown that the normal operational performance of a FeCrAl cladding can rival that of Zr-based alloys. Finally, it was found that optimizing the performance of FeCrAl cladding materials requires balancing the starting composition and microstructure in order to satisfy key performance indicators such as oxidation resistance and radiation tolerance. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗