FeCrAl Accident Tolerant Fuel Cladding: Insights from a Decade of Development
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Massey, Caleb↗
Engineering topics
Kane, Kenneth
Publications and source records attributed to Kane, Kenneth.
Development and Implementation of Digital Image Correlation for Out-of-Cell Burst Testing
This report details the progress made towards utilizing in-situ methods to observe and quantify unirradiated Zircaloy-4 (Zr4) fuel cladding deformation during out-of-cell design basis accident conditions, specifically during loss-of-coolant accident (LOCA) burst testing in the Severe Accident Test Station at Oak Ridge National Laboratory. Digital image correlation (DIC) was implemented to calculate cladding strain during burst and infrared (IR) thermography was used to attempt a correlation of strain with temperature and to investigate thermal gradients. For both techniques, early experimentation revealed experimental modifications were necessary for implementation. For DIC, this manifested as shortened cladding lengths which could be centered in front of the viewing chamber bored through the IR furnace side, and for IR thermography this meant conducting burst tests in air without a reaction tube and the incorporation of a SiC shell to eliminate direct IR reflections from the tungsten lamps. In the following, the technical issues and subsequent origin of these modifications are discussed, laying the framework for a proof-of-concept DIC/IR characterization of a burst test in the final section.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Air oxidation of yttrium hydride as a high temperature moderator for thermal neutron spectrum fission reactors
Yttrium hydride (YH x ) is an attractive moderator material for thermal neutron spectrum fission reactors requiring a small reactor core volume and has been selected as the neutron moderator for the Transformational Challenge Reactor (TCR), an advanced gas-cooled microreactor. Before YH x can be used in this application, it is important to understand the material response to off-normal conditions. In the present study, 550–650 °C isothermal dry air oxidation was performed to simulate a depressurized loss of force circulation (DLOFC) event. The oxidation was performed using thermogravimetric analysis (TGA) on bulk crack-free YHx coupons. Oxidation studies were also performed on Y coupons to elucidate the impact of H on oxidation. Both the chemistry and distribution of processing impurities were found to strongly affect oxidization behavior on a batch-to-batch basis. Regardless of batch, YHx oxidized at a significantly lower rate than Y at all temperatures, and the lower rate was directly correlated with increased hydride content. Metallic Y exhibited complex exponential kinetics, whereas YH x also exhibited complex kinetics but gained considerably less mass. According to literature reports on protonic and native-ion conductivities of Y 2 O 3 and mass spectrometry analysis of gaseous reaction products formed during the oxidation of YH x , a mechanism for the reduced oxidation rate of yttrium hydride is suggested.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FY21 burst activities with coated Zircaloy-4 under accident conditions
This report summarizes the results of Advanced Fuels Campaign (AFC) accident-tolerant fuel (ATF) burst activities. Nuclear service grade Zry-4 was procured and coated with a 7-micron thick Cr coating. The coating quality was investigated, and there were several defects at the Cr/Zry-4 interface due to the surface roughness of the as received Zry-4 tubing. To provide insight into the effect of coating defects on the cladding performance under accident scenarios, the unirradiated uncoated and coated material was tested under loss of coolant accident (LOCA) and pellet cladding mechanical interaction (PCMI) reactivity insertion accident (RIA) conditions. The defected coating appeared to have no impact on the cladding performance under these scenarios when compared to the as-received cladding material.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Quantifying High Temperature Corrosion
Alloys designed for high temperature service can form a variety of surface oxides or scales based on the alloy composition, service temperature, gas environment, surface deposits, etc. The most severe environments, including molten salts and liquid metals can result in significant metal loss, void formation and/or pitting. Standard and evolving practices are discussed for measuring these types of degradationincluding measuring oxide scale thickness, internal oxidation and/or metal loss, correlating them to mass change and reporting the results. Manual and automated methodologies to measure oxide thickness and metal loss are reviewed and compared in order to provide a summary of techniques and identify the best practices for quantifying a variety of materials and damage mechanisms and producing statistically meaningful results. Such techniques create datasets useful for improved understanding of corrosion degradation mechanisms, potentially better predictive models and enhanced data analytics.
Su, Yi Feng↗
Oxidation of 3D-printed SiC in air and steam environments
Here, the high-temperature oxidation of additively manufactured and chemically vapor infiltrated (3D-printed SiC) has been compared to chemical vapor deposited (CVD) SiC. 100-h isothermal exposures were conducted at 1425° and 1300°C at 1 atm under both dry air and steam environments. A SiC reaction tube was utilized to reduce silica volatility. After steam oxidation at 1425° and 1300°C, on the 3D-printed SiC surface, which was intrinsically rougher than the CVD surface, scales were 70%–90% thicker at the convex regions compared to concave/flat regions. In the convex regions, large cracks perpendicular to the oxidizing interface were observed. After dry air oxidation, scale thicknesses were comparable between 3D-printed SiC and CVD SiC, regardless of geometry. Finite element modeling, conducted to elucidate the relationship between SiC geometry and ß- to α-cristobalite transformation stress, determined cristobalite transformation tensile stresses to be on the order of 103 MPa during cool down, assuming a 6 vol% reduction. Compared to flat SiC substrates, tensile transformation stresses were elevated at concave regions and relaxed at convex regions. Combined with specimen mass gain (accounting for the rougher surface) of 3D-printed SiC being 15%–32% higher for 3D-printed SiC after 1300°C and 1425°C steam oxidation, the work presented concludes that the increased oxidation of 3D-printed SiC is primarily caused by tensile hoop stresses driven by oxidation volume expansion. Lastly, the efficacy of the 3D-printing method is demonstrated through the production of tristructural isotropic imbedded 3D-printed SiC fuel forms.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Oxidation of Yttrium Hydride Moderator for the Transformational Challenge Reactor
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Kane, Kenneth↗
Characterization of the Benefit of APS Flash Coatings in Improving TBC Lifetime
The addition of an air plasma sprayed (APS) “flash” layer on top of a high velocity oxygen fuel (HVOF) bond coating has been shown to extend the lifetime of thermal barrier coatings. A series of furnace cycle tests (FCTs) has been conducted at 1100 °C in air + 10% H 2 O to study the benefit of flash coatings on rod and disk alloy 247 specimens and provide a better mechanistic understanding of their benefit. Flash coatings of NiCoCrAlY and NiCoCrAlYHfSi both improved the FCT lifetime of rod specimens tested in 100-h cycles and disk specimens tested in 1-h cycles. In 1-h cycles, the NiCoCrAlY flash coating significantly outperformed an HVOF-only NiCoCrAlYHfSi bond coating and a NiCoCrAlYHfSi flash coating. Both flash coatings increased the bond coating roughness compared to HVOF. During exposure, the flash layer formed an intermixed alumina-metal layer that appeared to inhibit crack formation. Using a time series of observations, the lower Y + Hf content in the Y-only flash coating appeared to reduce Al consumption. The HVOF layer acted as a source of Al for the adjacent mixed zone. A second series of specimens included a fully APS bond coating where oxide had penetrated through the entire coating to the substrate after only 100, 1-h cycles and lifetime was similar to an HVOF-only bond coating. The inner HVOF layer with the outer APS flash coating prevented this complete penetration from occurring.
Pint, Bruce↗
Computational Methods to Accelerate Development of Corrosion Resistant Coatings for Industrial Gas Turbines
Oxidation resistant overlay coatings protect the underlying superalloy component in industrial gas turbines from oxidation attack. Rate of depletion of the Al-rich β-phase in the bond coat governs the lifetime of these coatings. The applicability of a computational method in accelerating the development of corrosion resistant coatings and significantly reducing the extensive experimental effort to predict coating lifetimes and microstructural changes in three-coated Ni-based superalloys for real operational durations (20–40 kh) was undertaken in the present study. Scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDX), and electron microprobe analysis (EPMA) were employed to characterize MCrAlY-coated superalloy substrates (1483, 247 and X4) after exposure at 900 °C in air + 10% H 2 O for up to 20,000 h. The model predicted the longest coating lifetime for the coating on X4 substrate. Precipitation of γ' in the coatings was correctly predicted for all three coating systems. Additionally, the model was able to predict the formation of topologically close packed (TCP)-phases in the investigated coating systems.
Pillai, Rishi R.↗
Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions
A novel experiment to simulate cyclic dryout in boiling water reactors has been developed to better understand the performance of nuclear grade FeCrAl cladding in a BWR during dryout conditions caused by an Anticipated Operational Occurrence or Anticipated Transient Without SCRAM - both of which are Design Basis Accidents. Internally pressurized C26 M FeCrAl alloy cladding and Zircaloy-2 cladding were subjected to rapid 300°-650 °C thermal cycling in a steam environment; actual maximum temperatures were found to vary between materials but were always above 650 °C. In the range of 32–55 MPa hoop stress, Zircaloy-2 cladding burst within 1–16 cycles (about 100 s of dryout duration above 600 °C), while at 76 MPa hoop stress, C26 M cladding remained virtually undeformed after completing 54 cycles (over 1000 s of dryout duration above 600 °C). Higher temperature 300°-700 °C and 300°C–800 °C cycling experiments had to be performed to induce C26 M burst – failure occurred after 20 cycles in the former and during the first cycle in the latter. Zircaloy-2 and C26 M failure criteria were used to generate hoop stress specific dryout lifetimes. Overall, the simulated cyclic dryout experiments show that nuclear grade C26 M cladding has significantly enhanced survivability under dryout conditions relative to Zircaloy-2.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗