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Ang, Caen K.

Publications and source records attributed to Ang, Caen K..

9 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding

In this work, high purity SiC and SiC/SiC composites coated with commercial TiN, Cr, CrN, or CrN/Cr multilayer coatings were irradiated in Ar or flowing PWR water in the Massachusetts Institute of Technology Nuclear Reactor Laboratory (MITR). Irradiation in Ar was performed in the core. In the water environment, identical samples were placed in one of three different locations: in-core, providing exposure to neutron damage and radiolysis-affected water; above-core, where samples were exposed to radiolysis-affected water but not neutron damage, or outside of the core, where samples were exposed to the coolant water without the effects of radiation. Radiation in Ar revealed significant cracking of all but the TiN coatings, attributed to differential swelling between the coating and substrate. Lattice swelling was not observed in any of the coatings, but 0.2% void swelling was observed in the Cr coating. All of the coatings failed during water exposures in the core. Likewise, CrN/Cr spalled in each condition. Cr was protective, except under radiation damage as a result of cracking, and TiN severely degraded in the core with no coating was found following exposure. A SiC/coating ATF cladding system is anticipated to perform adequately following improvements in coating ductility and purity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Preliminary characterization and projections of PVD coatings on SiC cladding for Light Water Reactors

Physical Vapor Deposition (PVD) based CrN, Cr and TiN are currently the leading mitigation coatings for SiC-based Light Water Reactor (LWR) fuel cladding. This manuscript describes the morphology, phase analysis and elemental composition of coatings deposited by commercial vendors on SiC clad materials. The coatings showed continuous crack-free morphology with few microstructural flaws. X-ray analysis and spectroscopy indicated high purity >99.9%. Mechanical tests of coating debonding (by ASTM D4541) and scratch-indentation (by ASTM E2546) were conducted. Strong interface bonding was observed and higher resistance to normal and shear stresses was observed from thicker coatings. Preliminary developments in scratch testing, phase analysis, future needs and expected behavior under irradiation are discussed.

Ang, Caen K.↗
DOE OSTI2018

Hydrothermal Corrosion of Coatings on Silicon Carbide in Boiling Water Reactor Conditions

SiC/SiC ceramic matrix composites are an attractive material for use as accident tolerant fuel cladding. However, despite its resistance to corrosion in high-temperature steam, SiC is known to dissolve in high-temperature water, necessitating the use of corrosion-resistant coatings. Here, samples were exposed in 288°C water with 2 ppmw dissolved oxygen to simulate boiling water reactor (BWR) normal water chemistry (NWC) and 288°C water with 150 ppbw dissolved hydrogen to simulate BWR hydrogen water chemistry. Several candidate coatings were tested: Cr, CrN, TiN, and ZrN. All coatings were applied by physical vapor deposition onto chemical vapor deposition SiC (CVD-SiC) substrates. Two uncoated SiC/SiC variants were also exposed and degraded more rapidly compared to CVD-SiC, especially in NWC. Cr and CrN coated specimens were found to have low mass change, suggesting they are promising candidate coatings. Lastly, TiN coatings were found to oxidize rapidly, suggesting that coatings would fully oxidize during service in BWR environments. ZrN coatings dissolved completely during exposure.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Response of isotopically tailored titanium diboride to neutron irradiation

The response of polycrystalline TiB 2 to neutron irradiation was investigated. The material was fabricated using isotopically enriched 11 B powders to minimize helium production via the 10 B(n, α) 7 Li reaction. Neutron irradiation was conducted at temperatures of ~200°C and ~600°C to a fast fluence of 2.4 × 10 25 n/m 2 (>0.1 MeV). The material exhibited some swelling, but less swelling at the higher irradiation temperature. No macroscopic damage was observed in the irradiated material, although moderate irradiation–induced micro–cracking was found in the irradiated TiB 2 . Furthermore, this study demonstrated improved radiation resistance of isotopically tailored TiB 2 compared with natural boron TiB 2 , which exhibited macroscopic fracture by irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Hydrothermal Corrosion of SiC Materials for Accident Tolerant Fuel Cladding with and Without Mitigation Coatings

As a candidate material for accident-tolerant fuel cladding for light water reactors (LWR), SiCf–SiC composite materials possess many attractive properties. However, prior work has shown that SiC is susceptible to aqueous dissolution in LWR coolant environments. To address this issue, candidate coatings have been developed to inhibit dissolution. For this study, CVD SiC samples were prepared with Cr, CrN, TiN, ZrN, NiCr, and Ni coatings. Uncoated SiC and SiCf–SiC samples were also prepared. The samples were exposed for 400 h in 288 °C water with 2 wppm DO in a constantly-refreshing autoclave to simulate BWR–NWC. Cr and Ni coated samples lost less mass than the uncoated SiC sample, indicating an improvement in performance. The CrN coating resisted oxidation, but some of the coating was lost due to poor adhesion. The TiN coated sample gained significant mass due to oxidation of the coating. ZrN and NiCr coatings showed significant corrosion attack. SiCf–SiC ceramic matrix composite materials dissolved much faster than the CVD SiC sample, demonstrating the need for mitigation coatings if CMCs are to be used in LWRs. This work demonstrates the promise of Cr, Ni and CrN coatings for corrosion mitigation in LWRs, and shows that NiCr and ZrN are not promising coating materials.

Raiman, Stephen S.↗
DOE OSTI2017

Evaluation of the First Generation Dual-purpose Coatings for SiC Cladding

The loss of coolant and station blackout at the Fukushima Daichi nuclear power station highlighted the need to improve the safety margins of current and future Light Water Reactors (LWRs). Accident Tolerant Fuel (ATF) development is designed to find solutions to mitigate or prevent such events through improvement of fuels and fuel cladding. One promising ATF concept is the replacement of zirconium-based alloy cladding with silicon carbide (SiC) fiber-reinforced SiC matrix (SiCf-SiC) composites, due to the inherent accident resistant features of SiC. However, during normal operations, SiC corrodes in reactor coolant potentially at an unacceptable rate. In addition, it is likely for the SiC-based cladding to undergo microcracking under an applied stress, which potentially degrades the hermeticity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Examination of Hybrid Metal Coatings for Mitigation of Fission Product Release and Corrosion Protection of LWR SiC/SiC

There is a need to increase the safety margins of current and future light water reactors (LWRs) due to the unfortunate events at Fukushima Daiichi Nuclear Plant. Safety is crucial to restore public confidence in nuclear energy, acknowledged as an economical, high-­density energy solution to climate change. The development of accident-­tolerant fuel (ATF) concepts is crucial to this endeavor. The objective of ATF is to delay the consequences of accident progression, being inset in high temperature steam and maintaining high thermomechanical strength for radionuclide retention. The use of advanced SiCf-­SiC composite as a substitute for zircaloy-­based cladding is being considered. However, at normal operations, SiC is vulnerable to the reactor coolant and may corrode at an unacceptable rate. As a ceramic-­matrix composite material, it is likely to undergo microcracking operation, which may compromise the ability to contain gaseous fission products. A proposed solution to both issues is the application of mitigation coatings for use in normal operations. At Oak Ridge National Laboratory (ORNL), three coating technologies have been investigated with industry collaborators and vendors. These are electrochemical deposition, cathodic arc physical vapor deposition (PVD hereafter) and vacuum plasma spray (VPS). The objective of this document is to summarize these processing technologies, the resultant as-­processed microstructures and properties of the coatings. In all processes, substrate constraint resulted in substantial tensile stresses within the coating layer. Each technology must mitigate this tensile stress. Electrochemical coatings use chromium as the coolant facing material, and are deposited on a nickel or carbon “bond coat”. This is economical but suffers microcracking in the chromium layer. PVD-­based coatings use chromium and titanium in both metallic form and nitrides, and can be deposited defense-­in-­depth as multilayers. This vapor method eliminated tensile stress during processing and coatings were up to ~30 μm thick without microcracking. VPS produced coatings based on Zircaloy-2, which has a proven reactor-­compatibility. The tensile stresses appearred to be partially mitigated by annealing. Analysis showed that VPS coatings required further optimizations to prevent adverse reactions with the substrate and need a minimum thickness of ~50 μm. In addition, development of coatings are constrained by neutronic depletion analysis, which clearly indicated enrichment as an issue if the coating is too thick. Based on the present work, the cathodic arc PVD technology was considered ready for the extensive testing and evaluation on cladding materials due to their ability to mitigate the excessive tensile stresses and the reasonable coating quality achieved. The VPS Zircaloy-­2 coating technology required additional development toward mitigation of issues related to the substrate reaction and porosity. In the future, PVD and VPS will have be scaled upon successful development and demonstration. Electrochemical coatings, which are proven scalability, currently require development to mitigate issues related to the tensile stress after deposition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes

Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge National Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗