Steam Oxidation of Zirconium-Yttrium Alloys from 500-1100 ;#8728; C 1100;#8728;C
Abstract not provided
Mouche, Peter↗
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Production of hydrogen by steam oxidation of calcium ferrite reduced with various coals
Production of H 2 by steam oxidation was demonstrated with calcium ferrite (CaFe 2 O 4 ) after coal gasification to produce CO with several types of coals. All coal types reacted with CaFe 2 O 4 to produce CO. The reduced CaFe 2 O 4 after the reaction with coal can also be oxidized with steam at 800 °C while producing of H 2 . Sub-bituminous coal and lignite coal had better reactivity with CaFe 2 O 4 than the bituminous coal. A ten-cycle test conducted with Wyodak coal/CaFe 2 O 4 at 800 °C showed stable production of H 2 during oxidation with steam. A 25-cycle steam oxidation test conducted after reduction of CaFe2O4 with lignite coal also showed a very stable performance. Steam oxidation is sufficient for maintaining the cyclic reactions. A larger scale (2 kg batch) production of CaFe 2 O 4 by wet granulation-tumbling method was successful. In conclusion, steam oxidation was demonstrated with the large-scale batch of CaFe 2 O 4 after reduction with Wyodak coal.
08 HYDROGEN↗
Study of the scale formed on super 304H boiler tube steels after long-term steam oxidation at high temperatures
Highlights: • Steam oxidation tests were performed on the conventional and shot-peened Super304H boiler tube. • A continuous amorphous SiO{sub 2} layer separated the metal from the scale in shot-peened specimen. • The SiO{sub 2} layer enhanced the formation of dense Cr{sub 2}O{sub 3} and FeCr{sub 2}O{sub 4} layers. • The scale on shot-peened specimens is stable upon steam temperature and time increase. Steam oxidation of a Super 304H boiler tube was performed at steam temperatures of 650 °C and 700 °C for time durations up to 5000 h. The oxidation scale outcomes which formed on as-received untreated and shot-peened (SP) tubes were studied using X-ray diffraction (XRD), scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), and transmission electron microscopy (TEM). γ-Grain nanocrystallization on the surface of the SP specimen generated an abundant fraction of grain boundary that guaranteed the formation of a continuous amorphous SiO{sub 2} layer at the metal-scale interface that enhanced the formation of dense Cr{sub 2}O{sub 3} and FeCr{sub 2}O{sub 4} in the scale. Scale breakaway along defects in the outer layer of the scale was accelerated by the steam temperature in the untreated specimens, whereas the scale that formed on the SP specimens appeared to be stable.
36 MATERIALS SCIENCE↗
Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering
In this study, a nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0wt. % at 1100°C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.
36 MATERIALS SCIENCE↗
Hydrogen Production by Steam Oxidation of Reduced CaFe 2 O 4 during Chemical Looping Coal Gasification: Equilibrium and Kinetic Analysis
CaFe 2 O 4 reduced during chemical looping coal gasification can be oxidized with steam for production of hydrogen. The conversion and equilibrium data of the oxidation of reduced CaFe 2 O 4 with steam were analyzed using steam concentrations of 10–25 vol % with a balance of helium and the temperature range of 1023–1123 K. Experimental data were evaluated to characterize H 2 production rates and oxidation isotherms. The film diffusion was the controlling step for the oxygen uptake process, and its parameters were determined on the basis of the uptake rate model. The steady-state oxygen uptake data were also evaluated with Langmuir, Temkin, Freundlich, and Dubinin–Radushkevich isotherms. The Langmuir isotherm model correlated the oxidation isotherms best, and the peak uptake capacity decreased with increasing temperatures, which suggests an exothermic process. In conclusion, the negative Gibbs free energy values (–Δ G °) implied the spontaneous nature of the uptake process at 1023–1123 K.
01 COAL, LIGNITE, AND PEAT↗
Enhanced steam oxidation resistance of uranium nitride nuclear fuel pellets
Here, the steam oxidation resistance of UN and UN-(20 vol%)ZrN fuel pellets is evaluated to enhance understanding of steam corrosion mechanisms in advanced nuclear fuel materials. In situ neutron diffraction shows the modified UN fuel pellets form a (U 0.77 ,Zr 0.23 )N solid-solution and the sole crystalline oxidation product detected in bulk is (U 0.77 ,Zr 0.23 )O 2 . U 2 N 3 is not detected in significant quantities during the steam oxidation of UN or (U 0.77 ,Zr 0.23 )N and stable lattice parameters show that hydriding does not take place. Steam oxidation rates, obtained via sequential Rietveld refinement show how (U 0.77 ,Zr 0.23 )N has a higher activation energy (79 ± 1 kJmol -1 vs. 50 ± 5 kJmol -1 ), higher onset temperature (430 °C vs. 400 °C) and slower reaction rates for steam oxidation up to 616 °C, than pure UN. Throughout, both UN and (U 0.77 ,Zr 0.23 )N exhibit linear (non-protective) oxidation kinetics, signifying that degradation of the fuel pellets is caused by the evolution of gaseous products at the interface followed by oxide scale spallation. This quantitative and mechanistic understanding of material degradation enables better defined operating regimes and points towards (U,Zr)N solid solutions as a promising strategy for the design of advanced nuclear fuel materials with enhanced steam corrosion resistance.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Multilayered Coatings for High-Temperature Steam Oxidation: TGA Studies up to 1000 °C
The aim of the study was to obtain a preliminary data focused on the steam oxidation performance of multilayered coatings deposited on 310 stainless steel and 800H Incoloy. The coatings were obtained through the aluminizing thermal diffusion technology with the formation of aluminides; some coating options had an additional top layer of selected composition. In this study, five different coating systems, as well as uncoated metals, were exposed at 1000 °C using thermogravimetric analysis instrumentation simulating an initial stage of steam oxidation corrosion. In this work, along with determination of the mass change of the samples, microanalyses using light optical microscopy, scanning electron microscopy (SEM) equipped with energy x-ray dispersive spectrometry (EDS), as well as surface micro-hardness determination, have been carried out. The designed coating systems demonstrated promising high-temperature steam oxidation performance with no breakaway degradation issues.
36 MATERIALS SCIENCE↗
Hydrothermal corrosion and steam oxidation behavior comparison of UAM and conventional Zry-4
Here, hydrothermal corrosion and steam exposures on ultrasonic additively manufactured (UAM) Zircaloy-4 (Zry-4) specimens were conducted to compare to wrought Zry-4 performance in both simulated pressurized water reactor (PWR) operation (330°C, 15 MPa, <50 ppb oxygen, 500h) and representative accident conditions (100% H 2 O(g) at 700, 900, and 1100°C for up to 1h). UAM Zry-4 specimens showed 50% higher mass gain in PWR conditions and 50-80% more mass gain in steam oxidation due to internal oxidation between build layers in both cases. UAM Zry-4 in 900°C steam showed breakaway oxidation leading to sample deformation. Titanium used as an interlayer in the UAM process was detected in the substrate after the 1100°C exposure. Additional experiments to better link as-built defects to oxidation behavior are needed.
36 MATERIALS SCIENCE↗
High temperature steam oxidation dynamics of U 3 Si 2 with alloying additions: Al, Cr, and Y
Uranium silicides are considered for advanced technology reactor fuels due to their enhanced thermal conductivity and high uranium density (U 3 Si and U 3 Si 2 ) compared to traditional UO 2 . Susceptibility to oxidation and wash out, in the event of a cladding breech, could limit the potential for deployment of silicides as accident tolerant fuels. Mitigating the water reaction for U 3 Si 2 could enable its use as an accident tolerant, high uranium density fuel or as a composite fuel constituent. Reported here is the impact of alloying additions of Al, Cr and Y on the high temperature, steam oxidation response of U 3 Si 2 . In addition to the thermogravimetric response, as melted microstructures, phase compositions and post oxidation analysis are also presented. Here, the investigation shows steam oxidation dynamics are altered, from non-alloyed U 3 Si 2 , under thermally ramped conditions. However, additional alloy development of these fuel forms is necessary for further consideration as candidate accident tolerant fuels in water-cooled reactor designs.
36 MATERIALS SCIENCE↗
Steam oxidation behavior of Ni-base superalloys 690, 725 and X-750 at 600 and 650°C
The 1-bar steam oxidation behavior of Ni-base superalloys 690, 725, and X-750 was evaluated at 600 and 650 °C for up to 5000 h. In addition to monitoring mass changes, microstructures of the 5000-h-exposed samples were characterized using scanning and transmission electron microscopy and energy dispersive x-ray spectroscopy. Together with thermodynamic calculations, the scale constitution and possible exfoliation was discussed. Alloy X-750 exhibited the best resistance to steam oxidation with the least metal thickness loss at 600 °C while alloy 690 at 650 °C. Alloy 725 exhibited more than two times metal thickness loss than the other two alloys at 650 °C.
36 MATERIALS SCIENCE↗
Steam oxidation and microstructural evolution of rare earth silicate environmental barrier coatings
A primary failure mode for environmental barrier coatings (EBCs) on SiC ceramic matrix composites (CMCs) is the oxidation of the intermediate Si-bond coating, where the formation of SiO 2 at the bond coating–EBC interface results in debonding and spallation. This work compares the microstructure evolution and steam oxidation kinetics of the Si-bond coating beneath yttrium/ytterbium disilicate ((Y/Yb)DS) and ytterbium disilicate/monosilicate (YbDS/YbMS) EBCs to better understand the impact of EBC composition on oxidation kinetics. After 500 1-h cycles at 1350°C, (Y/Yb)DS displayed a decreasing concentration of the monosilicate minor phase and increasing concentration of porosity as furnace cycling time increased, whereas the YbDS/YbMS EBC displayed negligible microstructural evolution. For both EBC systems, thermally grown oxide growth rates in steam were found to increase by approximately an order magnitude compared to dry air oxidation. The (Y/Yb)DS EBC displayed a reduced steam oxidation rate compared to YbDS/YbMS.
36 MATERIALS SCIENCE↗
Steam Oxidation and Burst Testing in the Severe Accident Test Station
Within six months of the March 2011 nuclear accident at Fukushima Daiichi, research began at Oak Ridge National Laboratory (ORNL) to simulate the high temperature steam oxidation accident conditions and identify potential accident tolerant fuel (ATF) cladding [1-14]. The ORNL laboratory equipment for steam oxidation equipment was expanded and made available to the entire ATF research community as the Severe Accident Test Station (SATS) in 2012 [5,1 5], including the capability to conduct the standard integral loss of coolant accident (LOCA) experiment [9]. A second SATS has now been deployed in an ORNL hot cell for handling irradiated specimens including reprocessed UO 2 /Zr fuel [16].
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A Phenomenological Durability Model for Environmental Barrier Coating Systems Subjected to Steam Oxidation
Environmental barrier coatings (EBC) are used to protect the silicon carbide fiber reinforced silicon carbide matrix (SiC/SiC) ceramic matrix composites (CMC) from rapid surface recession due to volatization by water vapor in a steam oxidation environment. A phenomenological modeling approach is outlined to estimate the durability of a given EBC system when exposed to steam oxidation. The report provides a step-by-step procedure to estimate the life of a given EBC system subjected to thermal loads only. Once calibrated, the methodology can predict the critical thickness of an oxide layer that will cause failure initiation/spallation as well as the remaining strength or stress that the EBC system can take for a given oxide layer thickness
Subodh K Mital↗
Preliminary results from Low Pressure Steam Oxidation Testing of ALD ZrN and ZrO2 Coating Deposited over UCN Fuel Kernels
Steam oxidation testing was used to investigate as-developed 100 nm ALD ZrN and 1000 nm ZrO 2 coatings deposited over UC1-xNx fuel kernels. The results of these tests were compared against oxidation of uncoated kernels. This work was performed to support development of coatings which both provide high temperature Zr metal diffusion barrier and also provide resistance against oxidation, especially against high temperature steam and/or air. This work was performed in collaboration with Oak Ridge National Laboratory (ORNL), using ORNL-provided samples. The oxidation tests yielded valuable information, including relative differences in performance of different coating materials and coating thicknesses.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
High-temperature steam oxidation study of irradiated FeCrAl defueled specimens
Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Metallurgical aspects influencing the resistance to steam oxidation and fracture toughness of select advanced replacement alloys for LWR core internals
Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Corrosion resistance in water environment is one of the fundamental properties required for LWR core internal materials. High-temperature steam oxidation tests are not only an accelerated life testing method to uncover potential failure modes in a short period of time, but also an approach to evaluate materials’ resistance to accidental scenarios. Coupons, prepared from fourteen candidate alloys that were selected under the ARRM program, were exposed to 1 bar full steam with ~10 part-per-billion oxygen content at 600 and 650°C for up to 5,000 h, which were weighed at 500-h intervals. Ni-base superalloys are an important class of alloys for reactor applications. Alloys 690, 725, and X- 750 were three of the selected Ni-base superalloys exposed to the steam tests. This report summarizes the metallurgical aspects, such as alloy thermodynamics and the nature of the oxides and the diffusivities of primary elements, and their effects on oxidation behavior of the three alloys. In addition to monitoring mass change, microstructures of the 5000h-exposed samples of alloys 690, 725, and X-750 were characterized using scanning and transmission electron microscopy and energy dispersive x-ray spectroscopy. Together with thermodynamic calculations, the scale constitution and its exfoliation behavior were discussed. Based on experimental and theoretical values, a new model was proposed to quantitatively estimate the metal thickness loss of the alloys. Alloy X-750 was found to have the best resistance to steam oxidation with the least metal thickness loss (<1 μm), which were slightly superior to alloy 690. Alloy 725 exhibited the worst resistance to steam oxidation with the greatest metal thickness loss, e.g., ~2.5 μm at 650°C, which is nearly three times of the metal thickness losses of alloys 690 and X-750 at 650°C. The greatest metal thickness loss of alloy 725 might be attributable to its high Mo content. Photos of the fracture toughness tested samples of eight alloys are presented, which will be correlated with their fracture toughness results and planned tensile test results to select representative alloy samples for detailed microstructural characterization to reveal their microstructure and properties relationships.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Correlations of the Steam Oxidation Rate Constant of BWR Alloy Zircaloy-2 at 800–1400 °C
Steam oxidation experiments were conducted at 800–1400 °C with boiling water reactor alloy Zircaloy-2 strip specimens. Sample weight gain measurements were performed on the oxidized specimens before and after the test and were compared to oxygen pickup calculations using the Cathcart–Pawel correlation. The results showed that Zircaloy-2 follows the parabolic law at temperatures above 1000 °C. At or below 1000 °C, the oxidation rate was very low when compared to Cathcart–Pawel correlation and can be represented by a cubic expression. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1000 °C and cubic rate constants are provided for temperatures at or below 1000 °C. Here, the weight gains calculated by our Arrhenius correlations are in excellent agreement with the measured sample weight gains at all test temperatures.
36 MATERIALS SCIENCE↗
High temperature steam oxidation of Cr-coated SiCf/SiC composite for LWR cladding applications
Chromium-coated silicon carbide fiber-reinforced ceramic matrix composites are potentially applicable to accident tolerant fuel cladding in light water reactors. Cr-coated silicon carbide composite, chemical vapor deposited silicon carbide, and pure Cr-metal were oxidized in 100% steam at atmospheric pressure for 4 hours using a thermogravimetric analyzer. Post-oxidation chemical analysis, phase identification and microstructural characterization indicates that the Cr coating layers did not adversely affect silicon carbide substrate under these conditions.
Lee, Yoonjo↗