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At least 37 records · Page 2

Effect of nickel on the oxidation behavior of FeCrAl alloy in simulated PWR and BWR conditions

FeCrAl alloys are being investigated as candidate materials for light water reactor fuel claddings. The chemical composition of FeCrAl determines its resistance to corrosion under normal operating conditions. The purpose of this paper is to investigate the effect of 0, 1, 3 wt% Ni addition on the passivation characteristics of Fe17Cr5.5Al in oxygenated and hydrogenated simulated reactor waters for 3-month immersion. Current results show that Ni addition improves the passivation of FeCrAl under oxygenated condition and decreases the mass loss under hydrogenated conditions.

36 MATERIALS SCIENCE↗

High-Temperature Creep Behavior of Thin-Walled FeCrAl Alloy Tubes

This report summarizes measured creep properties for FeCrAl alloy C26M measured using subsized specimens machined from prototypic thin-walled tubes. A representative tube fabricated in a previous large batch production effort was machined into dual-gauge axial specimens using electrical-discharge machining. Creep tests, comprising strain-rate jump tests (SRJ), jump stress (JS) and constant-stress (CS) tests, were conducted at temperatures ranging from 600-900ºC to determine creep mechanisms (elucidated from stress exponents) and activation energies to compare with prior data available in the literature at temperatures up to 650ºC. It was found that the results from the SRJ tests indicated that the highest stress exponent of 5.8 occurred at 600ºC, which decreased to the lowest stress exponent of 2.5 at 700ºC. At higher temperatures, the stress exponents increased slightly to a range of 2.9 to 4.3 between 750ºC and 900ºC. The high stress exponent value of 5.8 suggested dislocation glide creep mechanisms dominated creep deformation at 600ºC since high stress values were measured. The low stress exponent of 2.5 at 700ºC suggested diffusional creep mechanisms start to occur that aid in dislocation cross-slip and climb mechanisms. The slightly higher stress exponents of 2.9 to 3.9 may suggest different contributions of creep deformation mechanisms based on diffusional processes associated with Nabarro-Herring and Coble creep and dislocation creep that involve thermally activated cross-slip and climb mechanisms. These deformation mechanisms could have also been affected by the development of recovery and/or recrystallization processes in the microstructure of the C26M2 specimens. From the SJ and single CS tests, two sets of creep activation energies were calculated. The steady state creep values obtained from the SJ tests showed several abnormalities especially with the data at 850ºC. The comparison of the creep activation energies between the data from the SJ and CS tests showed higher creep rates for the SJ tests than the CS tests, indicating microstructure softening occurred in the C26M2 specimens during the SJ tests that led to the higher creep rates at stresses above the initial stress of 10 MPa. For the CS tests, the results showed steady increases in MCR with increasing temperature at each stress. The best linear fit to the MCR values at each stress was with data obtained at 800ºC, 850ºC and 900ºC. The lowest Q c value of 130.8 KJ/mole was observed with the highest stress of 25 MPa. As stress was lowered, the Q c values increased continuously to the highest value of 188.2 KJ/mole at 10 MPa. This trend suggests that the dominant creep mechanism is diffusional based dislocation climb since the decreasing activation energies with increasing stress promotes easier dislocation climb around obstacles. The comparison between the SJ and CS tests indicates that the best type of creep test performed on FeCrAl alloys such as C26M2 is the single CS test.

36 MATERIALS SCIENCE↗

Effect of surface oxides on tritium entrance and permeation in FeCrAl alloys for nuclear fuel cladding: a review

Abstract Iron-chromium-aluminum (FeCrAl) alloys are being considered for the cladding of uranium dioxide fuel in light water reactors (LWRs). FeCrAl alloys have good mechanical properties at temperatures of 300 °C and higher, and have superlative resistance to attack by steam at temperatures of up to 1000 °C and higher. A concern has been raised that the use of FeCrAl for cladding would result in a higher content of tritium in the reactor coolant as compared with the current system where the cladding is a zirconium based alloy. This review shows that the flux of tritium from the fuel rod cavities to the coolant across the fuel cladding wall will be greatly reduced by the presence of oxides on the surface of the cladding. The review of current literature and permeation data show that (a) protective oxides are expected to be present on both sides of the FeCrAl cladding, and (b) depending on the characteristics of these oxide layers it is reasonable to expect about two–three orders of magnitude reduction in tritium permeation, relative to the permeation response in clean, unoxidized condition for FeCrAl steels of interest, around 277 °C–377 °C temperatures.

Electrochemistry↗

Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding

As part of an effort to develop advanced fuel cladding for use in light water reactors, several FeCrAl alloys have been developed at Oak Ridge National Laboratory. A second generation of these alloys bearing additional alloying elements over the previously tested model alloys were tested in boiling water reactor (BWR) conditions to determine their resistance to hydrothermal corrosion. Coupons with alloy compositions Fe–10Cr–6Al–2Mo, Fe–13Cr–5Al–2Mo, Fe–13Cr–6Al–2Mo, Fe–13Cr–7Al–2Mo, and Fe–13Cr–5Al–2Mo–1Nb were tested in normal water chemistry (NWC) and hydrogen water chemistry (HWC) for 9 months in continuously refreshing autoclaves. Commercial FeCrAl alloy APMT and Zircaloy-2 were also tested for comparison. Among samples exposed to HWC, Fe–13Cr–7Al–2Mo performed worst with an average mass loss of 1.3 mg/cm2 over 9 months. This mass loss represents an estimated thickness loss of approximately 60 μm over 6 years. Samples exposed to NWC had very small mass losses of less than 0.15 mg/cm2 or mass gains up to 0.05 mg/cm2. Based on the results of this testing, the 2nd generation FeCrAl alloys tested exhibit low wall thickness loss and are suitable in terms of corrosion resistance for use as LWR cladding in BWR-HWC and BWR-NWC under normal operating conditions.

Raiman, Stephen↗

Solid State Additive Manufacturing of Oxide Dispersion Strengthened-FeCrAl Alloy Components for High-Temperature Supercritical CO2 Power Cycle Applications

In this research, material extrusion additive manufacturing (MEAM) has been explored as an SSAM method for fabrication of 3-D components using ODS-FeCrAl alloy. MEAM-printed 3-D parts go through a series of process steps, e.g., chemical treatment, thermal treatment etc. in order to create a high density fully metallic part. Final 3-D part quality is associated with various process steps, e.g., filament fabrication, print design, and thermal treatment optimization. In MEAM, a metal powder loaded filament is used for printing 3-D shapes. Composition of the filament is very important since it provides shape retention at low-to-medium temperature. There are a few commercially available metallic alloy filaments. However, FeCrAl alloy filaments are not commercially available. In this project, we came up with a method to fabricate high quality composite FeCrAl filaments that could be used for MEAM printing. Subsequently, we have optimized MEAM print methods to fabricate samples that are larger than 10 mm, by using a modular print design approach. Finally, we have been able to optimize the chemical and high temperature heat treatment steps to achieve very high density (>98%) in sintered MEAM-printed products.

36 MATERIALS SCIENCE↗

A data science approach for analysis and reconstruction of spinodal-like composition fields in irradiated FeCrAl alloys

A statistical method for the analysis of continuously distributed data representative of composition fluctuations in irradiated FeCrAl alloys acquired using Energy Dispersive X-ray Spectroscopy (EDS) method is presented. Using probability distribution functions, direct and cross-covariances between the elemental compositions, the effects of alloy composition and irradiation dose were investigated on the spatial distribution and length scale of composition fluctuations at the nanoscale. We have observed that, for neutron-irradiated FeCrAl alloys, the distribution of Fe and Cr followed a left-skewed and right-skewed distribution, respectively for all (average) alloy compositions and irradiation doses. The analysis also revealed enhanced spatial gradients in the elemental compositions at higher irradiation dose. Direct and cross-covariance estimates of the experimental data were also utilized for reconstruction of composition data through fitting it to a parametric form of the covariance functions. Linear Model of Coregionalization was used to determine the parameters of the covariance functions. Subsequently, a spectral method was utilized for simulating a realization of the alloy compositions. Close correspondence was observed between the experimental and the reconstructed data which was analyzed using probability distribution functions and covariance functions. Composition space of the experimental and reconstructed data and dislocation velocities as a function of applied stress and line directions over the entire composition maps were also examined.

36 MATERIALS SCIENCE↗

STEP Report on Advanced ODS FeCrAl Alloys for Fission Applications

Oak Ridge National Laboratory is developing advanced low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys for accident tolerant fuel (ATF) cladding. This report presents the characterization of two new ODS FeCrAl tubes by using powder commercially mechanically alloyed by Zoz GmbH, which underwent high-precision tube rolling with two different annealing schedules. This work shows that the ODS FeCrAl tube without recrystallization exhibits consistent mechanical behavior in comparison with a previously produced tube. The second tube, which underwent a full recrystallization step before the final tube rolling step, showed some circumferential cracking and exhibited limited ductility in the axial tube tensile tests. Based on current and prior work on the ODS FeCrAl alloy system, this report compares the current state of the scientific literature on the alloy class with that of wrought FeCrAl without oxide additions. Then, this report makes recommendations for future research directions that should be undertaken to increase the technology readiness level for ODS FeCrAl as a viable ATF candidate material. This report has been submitted as fulfillment of milestone M3FT-20OR020202064 titled, “STEP Report on Advanced ODS FeCrAl Alloys for Fission Applications” for the US Department of Energy Office of Nuclear Energy’s Advanced Fuel Campaign of the Fuel Cycle R&D program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hydrogen permeation in iron-chromium-aluminum (FeCrAl) alloys and the effects of microstructure and surface oxide

Iron–chromium–aluminum (FeCrAl) class alloys are candidates for use as cladding for accident-tolerant fuels and moderators. In this context, hydrogen isotope permeation in FeCrAl alloys is an important material property. Here, in the present work, the apparent permeability, effective diffusivity, and apparent solubility of hydrogen in the FeCrAl alloys C26M and Kanthal D (KD) were measured with gas-driven hydrogen permeation. Permeation measurements were conducted at temperatures of 400 to 700 °C and at gas-driven pressures from 1 to 100 kPa. In particular, the effect of grain size on hydrogen transport was studied with KD samples with three different microstructures: nanocrystalline (NC), ultra-fine grained (UFG), and coarse-grained (CG). The UFG and NC specimens had higher apparent activation energies (73.4 kJ mol -1 and 65.2 kJ mol -1 , respectively) for hydrogen permeability than the CG sample (46.9 kJ mol -1 ). An aluminum oxide layer formed on the primary- and secondary-side surfaces of all samples subjected to permeation experiments which demonstrated the propensity of FeCrAl alloys to form these innate oxide permeation barriers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Degradation and Failure Phenomena of Accident Tolerant Fuel Concepts: FeCrAl Alloy Cladding

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of accident tolerant fuel (ATF) in United States commercial nuclear power reactors. Pacific Northwest National Laboratory is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs. This report focuses specifically on the iron-chromium-aluminum (FeCrAl) alloys being investigated to replace zirconium-based alloys for fuel cladding and provides current state-of-the-industry information on material properties and fuel performance considerations under reactor operating conditions and design basis accident conditions. Currently, Global Nuclear Fuels (GNF) is the only U.S.-based fuel vendor considering FeCrAl cladding as a near-term ATF design. GNF has tested several different FeCrAl alloys including Kanthal APMT, C26M, and MA956; fueled C26M rods and unfueled Kanthal APMT, C26M, and MA956 rods have been inserted in lead test assemblies at Plants Hatch and Clinton. This report will focus on those three alloys specifically, with broader information given regarding in-reactor and ex-reactor testing of FeCrAl alloys. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of FeCrAl cladding concepts, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in light water reactors. Here, we have completed a first of its kind study where Al and Cr concentrations are varied systematically under neutron irradiation to elucidate the post irradiation microstructural and mechanical response as a direct result of alloying content. Neutron irradiations were performed on alloys with composition Fe-(10-13)Cr-(5-7)Al in wt.% at temperatures of 214, 357, and 557°C to a dose of ~1.8 dpa (displacements per atom). Dislocation loop sizes and dispersion characteristics did not show strong compositional dependence. However, there were noticeable effects of composition on the dispersion of Cr-rich α'-precipitates, particularly a decrease in number density with increasing Al content or a decrease in Cr content. The present results confirm previous studies indicating the Cr concentration in these precipitates is lower than that expected in binary FeCr alloys and that Al can act as a destabilizing alloying element for the deleterious α'-phase.

36 MATERIALS SCIENCE↗

Enhanced Mechanical Properties and Wear Resistance of FeCrAl Alloys at ~ 300°C and Higher Temperatures

Abstract Worldwide, light water reactors (LWRs) have been using zirconium (Zr)-based alloys for the cladding of the uranium dioxide fuel for more than 6 decades. Zr alloys oxidize rapidly in the presence of water and steam at temperatures > 450°C; therefore, they do not respond well to scenarios of loss of coolant accidents. There is a global effort by nuclear materials technologists to find more robust or stronger cladding materials for LWRs. One option is to use an iron-chromium-aluminum (FeCrAl) alloy since they have excellent resistance to high temperature oxidation and superior mechanical properties at LWR operation temperatures. Results show that (1) FeCrAl alloys have better mechanical properties than Zr alloy and are orders of magnitude more resistant to creep at temperatures higher than LWR normal operation conditions. (2) FeCrAl alloys have better resistance to fretting wear than Zr alloys at the normal operation conditions of LWRs.

Materials Science↗

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗

Corrosion behavior of ferritic FeCrAl alloys in simulated BWR normal water chemistry

Ferritic iron-chromium-aluminum (FeCrAl) alloys with varying Cr content were subjected to in-situ irradiation-corrosion with energetic protons or electron beams in pure water with various oxygen concentrations. Oxide microstructure and composition were characterized to compare the corrosion under irradiation with the thermal corrosion in the same water conditions. The dominant effect seems to be the production of H 2 O 2 at the surface by the intense proton and electron beams. The corrosion potential of FeCrAl under BWR-NWC condition is somewhat above the Cr 2 O 3 /CrO 4 2- phase boundary, oxidizing the Cr(III) to Cr(VI) in the outer part of the film, but with some Cr(III) remaining in the interior of the oxide. The beam increases the corrosion potential, significantly reducing Cr(III) content throughout the oxide. Finally, the crystallites on the surface, which form by re-precipitation from solution, do not contain Cr because at low potential Cr(III) is not soluble, and at high potential Cr(VI) is quite soluble.

36 MATERIALS SCIENCE↗

Investigation of deformation mechanisms in an advanced FeCrAl alloy using in-situ SEM-EBSD testing

The deformation mechanisms associated with uniaxial tensile testing are observed by conducting tensile experiments of an FeCrAl alloy using scanning electron microscopy (SEM) coupled with electron backscattered diffraction (EBSD). Prior to the deformation, investigated alloy was consisting grain and precipitate size of ~63.0 μm and ~6.7 μm, respectively. The recorded SEM micrographs and EBSD data at increasing levels of strains revealed the complex phenomena of slip bands’ formation in the presence of surface grain morphology evolution and their (001), (110) and (111) crystallographic planes distortions. The grains with orientation (110)||tensile direction (TD) shows higher shape change; however, (001)||TD and (111)||TD oriented grains show higher lattice gradient formation. Extracted information from the EBSD indicates that the crystallographic rotations drive towards specific, fiber-like texture in relation to the loading direction. Postmortem analysis of the recorded microstructure during the tensile deformation explains the phenomena of crack formation in the hard-intermetallic particles before the ultimate tensile strength (UTS). However, after the UTS, pores were identified in the neck that resulted from extensive plastic deformation. In-depth analysis was carried out to identify the cause of cracks and pores formation phenomena during the tensile test.

36 MATERIALS SCIENCE↗

Wrought FeCrAl alloy (C26M) cladding behavior and burst under simulated loss-of-coolant accident conditions

Cladding burst experiments for FeCrAl cladding were performed in the Severe Accident Test Station facility at Oak Ridge National Laboratory. These experiments were simulated using the BISON fuel performance code to better understand the cladding plastic behavior and failure under simulated loss-of-coolant accident conditions. 3D cladding surface boundary conditions were generated using composite axial and azimuthal profiles from experiment thermocouple data. To improve the simulation analysis capabilities in BISON for cladding burst behavior, new thermal creep, plasticity, and failure stress models specific to C26M, a wrought FeCrAl alloy, were developed and implemented. Initial cladding burst results indicated a general underprediction in the failure temperature of the six cladding burst simulations versus the observed failure temperatures. Close investigation of the experiment timing versus the underlying tensile test data revealed that, compared with the tensile specimens, the cladding tubes did not experience the same long holding time at high temperatures. New tensile tests were performed at high temperatures using a temperature ramp similar to the simulated loss-of-coolant accident experiments. These new tensile curves showed an approximately 80% increase in the ultimate tensile strength of the C26M alloy, indicating that a holding time of 10 min at 700 °C and 800 °C allows annealing to change the material microstructure. Using the updated tensile properties, the burst temperatures and stresses from the simulations showed remarkable agreement with the experimental results. This study was then extended by varying the initial pressure to highlight the burst temperature difference between standard Zircaloy-4 and C26M cladding under equivalent conditions. The results show that C26M has a burst temperature that is approximately 70–130 K greater than that of Zircaloy-4. In conclusion, these modeling predictions can be further improved by collecting high-temperature tensile data for C26M beyond the temperature ranges used in this work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Quantifying the resistance to dislocation glide in single phase FeCrAl alloy

The resistance to dislocation glide associated with slip systems {110}<111> and {112}<111> in single phase FeCrAl alloy is measured via micromechanical testing in a scanning electronic microscopy at room temperature. Two important factors, the shape and orientation of a pillar, are discussed with respect to the glide resistance and stress-strain response. Maximizing Schmid factor of one specific slip system while minimizing others is recommended in order to diminish obvious dislocations-induced hardening during in-situ testing. Apparent Schmid factor analysis is conducted to select grains with preferred orientations. Two types of pillars with conventional cylindrical shape or dog-bone shape are tested under compression to estimate the resistance to dislocation glide and evaluate the effect of pillar shape on the compression stress-strain response. One dog-bone pillar is tested under tension to check the tension-compression isotropy of dislocation slip. We discover that the shape of a pillar to a smaller extent affects the measured resistance but strongly influences the stress strain response. Cylindrical pillars exhibit apparent hardening associated with early yielding due to stress or strain concentration at contact region, while dog-bone pillars show an obvious yielding and continuous shearing without hardening. The resistance is 220 MPa for slip system {110}<111> and 230 MPa for slip system {112}<111>. Finite element analysis is performed to account for the influence of pillar shape and contact condition on mechanical response.

36 MATERIALS SCIENCE↗

Irradiation-induced amorphization of Fe-Y-based second phase particles in accident-tolerant FeCrAl alloys

Here, second phase intermetallic particles in an advanced accident-tolerant FeCrAl (Fe-13Cr-5Al-2Mo) alloy are formed in the α-Fe matrix during processing. These particles are prominently related to the added Y. Neutron irradiation to ~7 displacements per atom (dpa) with a dose rate of ~8.16 × 10 -7 dpa/s at 282 °C resulted in the amorphization of these precipitates which could degrade the mechanical properties of the FeCrAl alloys. Analytical electron microscopy and diffraction analysis combined with structural freedom analysis have been used to investigate the radiation resistance of the second phase particles. Radiation tolerance is closely linked to the particle Fe-Y content and can be tailored using the structure freedom value.

36 MATERIALS SCIENCE↗

Radiation Damage Mitigation in FeCrAl Alloy at Sub-Recrystallization Temperatures

Traditional defect recovery methods rely on high-temperature annealing, often exceeding 750 °C for FeCrAl. In this study, we introduce electron wind force (EWF)-assisted annealing as an alternative approach to mitigate irradiation-induced defects at significantly lower temperatures. FeCrAl samples irradiated with 5 MeV Zr 2+ ions at a dose of 10 14 cm −2 were annealed using EWF at 250 °C for 60 s. We demonstrate a remarkable transformation in the irradiated microstructure, where significant increases in kernel average misorientation (KAM) and low-angle grain boundaries (LAGBs) typically indicate heightened defect density; the use of EWF annealing reversed these effects. X-ray diffraction (XRD) confirmed these findings, showing substantial reductions in full width at half maximum (FWHM) values and a realignment of peak positions toward their original states, indicative of stress and defect recovery. To compare the effectiveness of EWF, we also conducted traditional thermal annealing at 250 °C for 7 h, which proved less effective in defect recovery as evidenced by less pronounced improvements in XRD FWHM values.

FeCrAl alloys↗