Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding
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73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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Behavior under LOCA conditions on enhanced accident tolerant chromium coated zircaloy-4 claddings
For enhanced accident tolerant fuels for light water reactors application, chromium coatings on zirconium based nuclear fuel claddings are developed and studied at CEA in the framework of the French CEA-EDF-AREVA collaborative program. The results obtained so far, mainly on Zircaloy-4 substrate, show very good corrosion resistance in nominal conditions and significant enhancement of the resistance of the material to oxidation in steam at high temperature (HT), up to 1300 Celsius degrees, with a drastic decrease of hydrogen release and/or pick-up. The present paper reports some new results obtained on chromium coated Zircaloy-4 claddings tested in loss-of-coolant accident (LOCA) conditions. In order to investigate the potential effect of the coating on the cladding mechanical behavior at HT and the capacity of the coating to sustain significant substrate deformation (i.e., during ballooning until burst occurrence) without generalized cracking/peeling, a preliminary limited set of internal pressure creep and temperature ramp tests have been performed in steam environment thanks to the EDGAR facility. The thermal-mechanical tests were done for testing/burst temperatures ranging from 600 C. degrees (α{sub Zr} phase domain) up to 1000 C. degrees (β{sub Zr} phase domain) on 50 cm long low-tin Zircaloy-4 cladding samples with a 15 μm thick outer chromium coating. It is shown that: -) whatever the applied temperature/pressure values, the chromium coating is still fully adherent after having experienced ballooning and burst, including at the vicinity of the burst opening where the Zircaloy-4 clad substrate is highly deformed; -) a HT strengthening effect of the coating on the overall creep clad behavior is evidenced when compared to uncoated Zircaloy-4 cladding materials tested in the same conditions; -) as a consequence, it is observed that, in the 600-750 C. degrees temperature range (α{sub Zr} phase domain) and after burst occurrence, the balloon sizes (i.e., 'uniform' and maximum hoop strains) are generally reduced when compared to uncoated materials; -) regarding the burst mechanism in the β{sub Zr} phase temperature range (1000 C. degrees), it is interesting to observe that, even if some ballooning occurred prior to the cladding failure, the actual burst openings are generally very small (in the order of 1 mm{sub 2} or less), reducing the risk of fuel fragments dispersal in the coolant.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The effect of temperature and burnup on U-10Zr metallic fuel chemical interaction with HT-9: A SEM-EDS study
The fuel cladding chemical interaction (FCCI) between Uranium-Zirconium-based metallic fuel and cladding materials during in-pile service is one of the most constraining phenomena affecting the performance of this fuel system. In this study, we investigated the effect of temperature and burnup on the FCCI development in two U-10 wt.% Zr (U-10Zr) fuel samples with HT-9 cladding irradiated as part of the MFF-3 irradiation test in the Fast Flux Test Facility (FFTF). One sample achieved a burnup of 13.1 at.% and operated with an average inner cladding temperature of 530°C, while the other achieved a burnup of 8.5 at.% and was subjected to an average inner cladding temperature of 615°C. Automated scanning electron microscopy (SEM) back-scattered electron (BSE) imaging of entire fuel cross-sections and SEM energy dispersive x-ray spectroscopy (EDS) analysis on specific fuel-cladding interface regions successfully provided a comprehensive characterization of the depth and type of interaction happening under different irradiation conditions. Further, our analysis shows that FCCI development on both fuel and cladding side is strongly influenced by the inner cladding temperature and, to some extent, the formation and integrity of Zr-rich layers between the fuel and cladding, while the impact of burnup and power is negligible. Measured FCCI thicknesses were compared to BISON simulations using both an empirical model based upon legacy data from the Experimental Breeder Reactor II (EBR II) irradiations and a mechanistic model currently under development for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, showing satisfactory agreement. Nonetheless, this comparison supports the need for additional microstructural characterization in intermediate ranges of temperature, power, and burnups in prototypic-length pins.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Comparison of ring compression testing to three point bend testing for unirradiated ZIRLO cladding
Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects. Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal history) and b) for critical strains in post drying handling, storage and accident conditions the 3 point bend strain tolerance is less affected by radial hydrides than the conventional ring compression test (the radial hydride related Quasi DBTT associated with a three point bend straining is lower (better) than that measured by the ring compression tests).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A study on texture stability and the biaxial creep behavior of as-hydrided CWSR Zircaloy-4 cladding at the effective stresses from 55 MPa to 65 MPa and temperatures from 300°C to 400°C
The creep rupture of high burnup used nuclear fuel (> 45 GWD/MTU) cladding is regarded as one of the failure mechanisms during long-term dry storage. A high amount of zirconium hydride in the cladding matrix would degrade the mechanical properties of the cladding especially leading to delayed hydride cracking. To better understand the influence of zirconium hydride on the biaxial thermal creep performance and the crystalline texture stability of Zircaloy-4 cladding, the pressurized tube technique is applied to test the durability of the as-hydrided material. Tests were performed on as-received Zircaloy-4 tubular specimens as well as-hydrided ones with targeted 300 wt parts per million (wppm) or 750 wppm hydrogen. The biaxial creep experiments were conducted at temperatures from 300 degrees C to 400 degrees C and at equivalent stresses at mid-wall from 55 MPa to 65 MPa. The hydridation process prior to creep tests induces the formation of FCC delta-hydride platelets along the circumferential direction of the tube. This alignment and phase structure of hydride show no significant change after biaxial creep tests. The creep strain-rate negatively depends on the hydrogen content. The synchrotron wide-angle X-ray diffraction (WAXD) technique and electron backscatter diffraction (EBSD) analysis were applied for the study of the crystallographic orientation relationship. zirconium-to-hydride grain orientation follows Shoji-Nishiyama crystallographic relationship. This relationship is stable before and after creep deformation. These results of creep performance and texture stability of Zircaloy-4 claddings can help support the design basis of interim and long-term dry storage facilities.
creep↗
Cold spray technology in nuclear energy applications: A review of recent advances
Cold spray technology is a promising solid-state, powder-based deposition methods for fabrication of coatings, near-net-shape manufacturing, and component repair. In recent years, this technology has been investigated for many applications in the nuclear energy sector. For example, it has been explored for the deposition of coatings of corrosion and oxidation materials on zirconium-alloys fuel claddings in light water reactors (LWR) for achieving more accident-tolerant fuel (ATF) cladding. Other examples of its use for nuclear energy applications include coatings for enhancements in wear resistance and heat transfer, near-net shape manufacturing of next-generation fuel claddings of oxide dispersion-strengthened (ODS) steels, and for mitigation and repair of potential chloride induced stress corrosion cracking in used fuel storage canister systems. This work reviews published literature on the application of cold spray process for nuclear energy applications, while pointing to gaps that have yet to be addressed.
36 MATERIALS SCIENCE↗
Multimodal analysis and characterization of the boehmite layer formed on AA6061 before and after alkaline etching
Low-enriched uranium (LEU) alloyed with 10% Mo (U-10Mo) is being considered as a promising alternative to oxide-based dispersion fuel with high-enriched uranium for use in research reactors. The configuration of this proposed LEU monolithic LEU fuel plate consists of a U-10Mo plate-type fuel foil with a 25 µm zirconium interlayer barrier clad with an aluminum alloy (AA6061). In certain research reactors, the clad AA6061 is coated with a boehmite layer to prevent corrosion. The boehmite layer has a high-pH passivation range, which makes it resistant to oxidation. Boehmite is usually formed on the AA6061 surface by autoclave processing. Before the boehmite layer is added, the surface of the AA6061 is cleaned using techniques such as polishing and wet etching. In this study, we use multimodal analysis to examine how pretreatment of AA6061 using polishing followed by alkaline etching affects the chemical composition of the boehmite layer. X-ray photoelectron microscopy (XPS), transmission electron microscopy (TEM), and x-ray diffraction (XRD) were used to study the chemical changes in the boehmite layer caused by alkaline etching pretreatment. XPS provides quantitative analysis for the Al:O ratio as well as oxidation states present on the surface, which suggests slight oxidation of the boehmite surface after alkaline etching of the AA6061 surface. We further explored this suggested oxidation of the boehmite surface using high-resolution transmission electron microscopy with selected area electron diffraction (SAED) and grazing incidence x-ray diffraction (GI-XRD), which suggested only a small amount of aluminum oxide at the surface. The multimodal analysis and imaging yielded new insights for optimizing boehmite growth on AA6061 for research reactors.
36 MATERIALS SCIENCE↗
THE EFFECTS OF FABRICATION INDUCED RESIDUAL STRESS-STRAIN STATES ON THE IRRADIATION PERFORMANCE OF MONOLITHIC MINI-PLATES
This article evaluates the initial stress-strain state of the monolithic fuel plates and their effects on the overall irradiation performance. The monolithic fuel is a proposed fuel form for high-performance research reactors to achieve higher uranium densities within the core to allow the use of low-enriched uranium fuel. These plate-type fuels consist of a high density low enrichment Uranium-Molybdenum alloy based fuel, a diffusion barrier made of Zirconium, and a cladding made of Aluminum. Because the fabrication of monolithic plates involves multiple stages and the foil material is subjected to various thermo-mechanical processes, a presence of significant stress-strain fields is likely. In addition, the choice of the fabrication process may also cause variations in the foil’s microstructure. There are interests if the overall irradiation performance could be affected by the different initial stress-strain state and the microstructural characteristics. To investigate these concerns, a representative plate was simulated for various initial stress-strain states. First, the co-rolling process for thickness reduction stage of the foil was simulated. For this, a scaled version of the process was evaluated to identify distinct stress profiles on the selected paths. These profiles were then used to define an initial state for the foil before Hot Isostatic Pressing. Additional HIP simulations were also performed to study the cases with stress-free foils prior HIP bonding. The purpose of these simulations was to mimic the stress relief treatment after co-rolling, which aims to remove the residual stresses and minimize the material anisotropy. For the HIP process with initially stress-free foil, several bonding temperatures were considered, including the nominal bonding temperature. The final case considered a plate without initial stresses and strains, assuming a hypothetical case in which the plate would be subjected to a post-fabrication heat treatment to relieve the process-induced stresses before irradiation. Finally, the irradiation processes were simulated for all the cases with distinct pre-irradiation stress-strain profiles. For all cases, the same irradiation parameters were used. The resulting distortions, stress-strain fields and temperature profiles were gathered on selected locations. Finally, a comparative evaluation was made to determine the sensitivity to the pre-irradiation stress-strain variations. The irradiation simulations have revealed that fabrication induced stresses in fuel would be relieved quickly due to large irradiation creep strain. The results indicated that the stress-strain behavior of the fuel foil is not affected significantly by the fabrication process. It is important to note that this excludes microstructural characteristics, which may have important implications on overall behavior. In the cladding however, the stresses continue increase and additional plastic strains are generated. This is due to irradiation hardening by fast neutrons. Additional stresses and strains are formed in the cladding as a result of displacements caused by the volumetric swelling of the fuel foil.
42 ENGINEERING↗
Fabrication, Welding, and Inspection Techniques in Preparation for In-Reactor Testing of Accident Tolerant Fuels
After the Fukushima events in 2011, DOE-NE in collaboration with nuclear industry shifted R&D emphasis to accident performance of LWR fuels under extended loss of active cooling and steam exposure. DOE-NE has created a roadmap for the “Development of Light Water Reactor Fuels with Enhanced Accident Tolerance.” The mission of the Accident Tolerant Fuel (ATF) Roadmap is to develop the next generation of LWR fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. The ultimate goal of the ATF roadmap is to support the insertion of lead fuel rods (LFRs) or lead fuel assemblies (LFAs) of an Accident Tolerant Fuel into a commercial LWR within 10 years (i.e., by the end of FY-2022). As a step toward this goal, an irradiation test series has been developed to assess the performance of proposed ATF concepts under normal LWR operating conditions. Data generated by this test program will be used to establish the feasibility of certain aspects of proposed ATF concepts, as well as provide information to support screening among concepts; as such, it is an integral part of Phase I: Feasibility Assessment and Down-Selection outlined in the ATF Roadmap. This irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL), and it has been designated as the ATF-1 test series. Current fission reactors use zirconium-based fuel cladding because of its extremely low macroscopic thermal neutron absorption cross-section, good high temperature strength, and decent corrosion resistance. However, advanced, innovative materials may provide these same benefits while increasing reactor safety margin, core power density, and fuel utilization. These advanced fuel cladding systems will allow revolutionary cladding performance and enhanced fuel mechanical designs, however, challenges exist in design, analysis and fabrication of innovative, never before tested, fuel cladding systems for in-reactor testing. This paper highlights the challenges associated with design, fabrication and welding, and inspection of innovative materials and actions taken to address those challenges in preparation for the Phase I ATR irradiation testing. The lessons learned from Phase I of this experiment can be used to guide researchers for design and analysis of future in-reactor testing of advanced fuel cladding systems.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding
We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloy reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Mechanical behavior at high temperatures of highly oxygen- or hydrogen-enriched α and prior-β phases of Zirconium alloys - 2016-0063
During a hypothetical loss-of-coolant accident, zirconium alloy fuel claddings can be loaded by internal pressure and exposed to steam at high temperatures (HTs) up to 1,200 deg. C and then cooled and quenched in water. A significant fraction of the oxygen that reacts with the cladding during HT oxidation diffuses beneath the oxide through the metallic substrate. This induces a progressive transformation of the metallic β{sub Zr}-phase layer into an intermediate layer of α{sub Zr}(O) phase containing up to 7 weight % (wt.%) oxygen. Furthermore, in some specific conditions, the cladding may rapidly absorb a significant amount of hydrogen during steam exposition at HTs. As a β{sub Zr} stabilizer, hydrogen would mainly diffuse and concentrate up to several thousands of parts per million by weight (wppm) into the inner β{sub Zr}-phase layer. This study intends to provide new and more comprehensive data on the HT mechanical behavior of the α{sub Zr}(O) and (prior-)β{sub Zr} phases containing high contents of oxygen and hydrogen, respectively. Model samples produced from M5{sup R} and Zircaloy-4 cladding tubes homogeneously charged in oxygen (less than or equal to 6 wt.%) and hydrogen (less than or equal to 3,000 wppm), respectively, were prepared. Their mechanical behavior was determined under vacuum between 800 and 1,100 deg. C for the oxygen-enriched α{sub Zr} phase and in air between 700 and 20 deg. C after cooling from the β{sub Zr} temperature domain for the hydrogen-enriched (prior-)β{sub Zr} phase. The α{sub Zr} phase was substantially strengthened and embrittled by oxygen. The contribution of the α{sub Zr}(O) layer to the HT creep behavior of an oxidized fuel cladding tube subjected to internal pressure is evaluated by finite element analysis. Mechanical strength and ductility of the model (prior-)β{sub Zr} phase appear to be affected by hydrogen contents of 2000-3000 wppm in ways that depend on temperature. (authors)
36 MATERIALS SCIENCE↗
TEV-3776 Hybrid – ZIRCEX Process Demonstration Siting Study
The purpose of this technical evaluation (TEV) is to detail the siting considerations for supporting pre-conceptual design of the engineering-scale Hybrid – ZIRCEX Process (formerly known as Hybrid Uranium Recovery Demonstration [HURD]) demonstration. This process represents the first-of a kind integration of a select group of technologies meant to recover high enriched uranium (HEU) from zirconium and aluminum clad used nuclear fuel (UNF) for down-blending to produce high assay low enriched uranium (HALEU).
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Al/Zr/U-10Mo Foil Sensitivity Analyses
This report describes modeling work performed to analyze typical uranium (U)-10 wt percent molybdenum (Mo) foils co-rolled with zirconium (Zr) and clad in aluminum (Al). The purpose of the analyses was to predict foil integrity, while accounting for the mechanical properties, thickness, and roughness of the Zr-layer coating.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A LOCA Analysis Tool: Coupling RELAP5-3D to BISON
Experimental evidence illustrates that at burnups slightly above the current regulatory limit of a rod-averaged burnup of 62 MWd/kgU, the ceramic UO 2 inside light-water reactor fuel rods becomes susceptible to a phenomenon known as fuel fragmentation, relocation, and dispersal (FFRD) during a loss of coolant accident (LOCA) transient. The severity of FFRD is strongly influenced by the zirconium-based (Zircaloy) cladding behavior during the LOCA transient. A Technology Commercialization Fund (TCF) project was awarded to an Electric Power Research Institute (EPRI)/Idaho National Laboratory team to create a LOCA analysis tool that couples BISON to the systems/thermal-hydraulics code RELAP5-3D [1] for analysis of LOCA scenarios. In addition, further refinements to existing BISON models were identified as necessary to more accurately represent more recent experimental evidence from the Studsvik Cladding Integrity Project (SCIP) and other experimental programs.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Effects of thermal treatment on the co-rolled U-Mo fuel foils
A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface between the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Mechanical testing of hydrided fuel cladding
During light water reactor operations, hydrogen is generated from the oxidation of zirconium alloy materials. This hydrogen can be absorbed into the cladding and form zirconium hydrides. At high temperatures and hoop stresses, these normally circumferential oriented hydrides in high burnup used nuclear fuel (HB-UNF) may reorient in the radial direction and affect the mechanical properties of the cladding. Understanding these changes in mechanical properties will aid in analyzing how UNF will perform during storage and transportation. Systems, presented in the article, have been designed to introduce hydrides into cladding specimens, to reorient cladding hydrides, and to test and measure the mechanical properties of these specimens. The hydriding systems consist of a tube furnace capable of pre-hydriding cladding specimens and reorienting cladding hydrides and surface conditioning equipment. Mechanical testing (i.e., tensile, compression, and burst testing) will be performed on both pre-hydrided cladding and cladding with reoriented hydrides, and the results will be compared with those of as-manufactured cladding. Tensile and compression testing will be performed using a 5 kilo-pound load frame; burst testing will be performed on a purpose-built test system. Strain is measured with a digital image correlation (DIC) system. DIC is a real time optical measurement technique that utilizes high-definition imagery to measure object deformation. It is planned that tensile, compression, and burst testing of cladding with reoriented hydrides will begin this year.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Design of Laboratory System to Recover Tritium from Zirconium Recycle and Plan to Test System
The objective of the tritium separations task is to develop a viable process to separate tritium from the zirconium recycle off-gas stream. In support of this long-term objective, the goal of the present work is to design a test system and prepare a test plan for an alternate tritium capture method suitable for, and compatible with, recycle of the reactive gas (i.e., chlorine) in the zirconium recycle process. The zirconium recycle process is under development to recover zirconium from the Zircaloy® cladding component of used nuclear fuel. Small quantities of all isotopes of hydrogen (all H) are trapped in the Zircaloy as zirconium hydride. Processing releases and converts this hydrogen to hydrogen chloride, including tritium-substitu ted hydrogen chloride, 3 HCl (a.k.a., TCl). This compound of tritium must be removed from the zirconium recycle process off-gas and sequestered to satisfy regulations governing emissions of tritium.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Fuel Performance Simulation of Iron-Chrome-Aluminum Cladding during Steady-State LWR Operation
During a severe accident scenario, such as a loss-of-coolant accident, a nuclear reactor may lose its capacity to cool its fuel. The fuel rod temperatures begin to increases, quickly reaching temperatures where the coolant begins to evaporate. This lowers the coolant level in the core and eventually uncovers the fuel. Without sufficient heat transfer from the fuel rods to the coolant, the temperature of the fuel and cladding will increase dramatically. In the case of traditional zirconium-based alloys, as the temperature of the cladding reaches ∼1200 deg. C the zirconium rapidly begins to oxidize with H{sub 2}O in the coolant and steam. This oxidation reaction causes both thinning of the Zr-alloy cladding as the metal reacts and releases large amounts of H{sub 2} gas into the reactor pressure vessel. In order to increase the safety margin of LWR fuel in severe accident scenarios, several alternative cladding materials have been proposed to replace the currently used zirconium-based alloys. Of these materials, there is a particular focus on select iron-chrome-aluminum (FeCrAl) alloys because of slower oxidation Kinetics in high temperature steam than zirconium-based alloys. This increased oxidation resistance may provide additional coping time to mitigate any further damage resulting from an accident. However, by changing the cladding material, the neutronic characteristics of the core will change. FeCrAl alloys exhibit thermal neutron absorption cross sections nearly ∼12-16 x greater than Zr-alloys. This decreases the reactivity of the core, subsequently decreasing the potential cycle length for operation. To counteract this, fissile material may be added in the form of increased enrichment or increased fuel mass or the cladding thickness might be decreased, effectively reducing the neutronic penalty from using the alloy. In this study, all three changes are considered. To assess the viability of FeCrAl as an alternative cladding to Zr-alloys, engineering-scale fuel performance analysis is used to monitor the cladding and integral fuel rod behavior through its operation. Ongoing work is being performed to identify key parameters essential to modeling FeCrAl alloys for LWR applications and properly model fuel compliance and stress relief mechanisms for UO{sub 2} fuel. To improve the predictive capability of fuel performance codes for FeCrAl cladding under normal operation, high-fidelity mechanistic/physics-based material models are needed. This information can be generated and improved by access to experimental data on targeted FeCrAl alloys. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗