• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.597-609
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• 2020

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, Cr_xN_y layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360℃ in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400℃ related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

• Applied Microscopy
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• v.52
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• pp.5.1-5.10
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• 2022

The use of Pipelines for long-distance transportation of crude oil, natural gas and similar applications is increasing and has pivotal importance in recent times. High specific strength plays a crucial role in improving transport efficiency through increased pressure and improved laying efficiency through reduced diameter and weight of line pipes. TRIP-based high-strength and high-ductility alloys comprise a mixture of ferrite, bainite, and retained austenite that provide excellent mechanical properties such as dimensional stability, fatigue strength, and impact toughness. This study performs microstructure analysis using both Nital etching and LePera etching methods. At the time of Nital etching, it is difficult to distinctly observe second phase. However, using LePera etching conditions it is possible to distinctly measure the M/A phase and ferrite matrix. The fraction measurement was done using OM and SEM images which give similar results for the average volume fraction of the phases. Although it is possible to distinguish the M/A phase from the SEM image of the sample subjected to LePera etching. However, using Nital etching is nearly impossible. Nital etching is good at specific phase analysis than LePera etching when using SEM images.

• Corrosion Science and Technology
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• v.22 no.5
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• pp.359-367
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• 2023

The localized corrosion resistance of the Ni-based Inconel 718 alloy after solution heat treatment was evaluated using electrochemical techniques in a solution of 25 wt% NaCl and 0.5 wt% acetic acid. Solution heat treatment at 1050 ℃ for 2.5 hours resulted in an increased average grain diameter. Both Ti carbides (10 ㎛ diameter) and Nb-Mo carbides (1 - 9 ㎛ diameter) were distributed throughout the material. Despite heat treatment, the shape and composition of these carbides remained consistent. An increase in solution temperature led to a decrease in pitting potential value. However, the pitting potential value of solution heat-treated Inconel 718 was consistently higher than that of as-received Inconel 718 at all tested temperatures. Localized corrosion initiation occurred at 0.4 V_SSE in a temperature environment of 80 ℃ for both as-received and solution heat-treated Inconel 718 alloys. X-ray photoelectron spectroscopic analysis indicated that the composition of the passive film formed on specimen surfaces remained largely unchanged after solution heat treatment, with O1s, Cr2p_3/2, Fe2p_3/2, and Ni2p_3/2 present. The difference in localized corrosion resistance between as-received and solution heat-treated Inconel 718 alloys was attributable to microstructural changes induced by the heat treatment process.

• Tribology and Lubricants
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• v.40 no.3
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• pp.97-102
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• 2024

The next-generation Molten Salt Reactor is known for its high safety because it uses nuclear fuel dissolved in high-temperature molten salt, unlike traditional solid atomic fuel methods. However, the high-temperature molten salt causes severe corrosion in internal structural materials, threatening the reactor's safety. Therefore, it is crucial to investigate the high-temperature corrosion resistance and wear performance of materials used in reactors to ensure safety. In this study, the high-temperature corrosion resistances and wear performances of corrosion samples in a NaCl-MgCl2-KCl (20-40-40 [wt%]) molten salt are investigated to evaluate the applicability of economically viable stainless steels, 316SS and 304SS. Hastelloy C276 and a new alloy containing a small amount of Nb are used as reference samples for comparative analysis. The mass loss, mass loss rate per unit volume, and surface roughness of each sample are measured to understand the corrosion mechanisms. Scanning electron microscopy and energy-dispersive spectroscopy analyses are employed to analyze the corrosion mechanisms. Wear tests on the corroded samples are also conducted to assess the extent of corrosion. Based on the experimental results, we predict the lifespans of the materials and evaluate their suitability as candidate materials for molten salt reactors. The data obtained from the experiments provide a valuable database for structural materials that can enhance the stability of molten salt reactors and recommend high-temperature corrosion-resistant materials suitable for next-generation reactors.

• Journal of the Korean Society for Nondestructive Testing
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• v.34 no.3
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• pp.254-259
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• 2014

A pressurized heavy water reactor (PHWR) core has 380 fuel channels contained and supported by a horizontal cylindrical vessel known as the calandria, whereas a pressurized water reactor (PWR) has only a single reactor vessel. The pressure tube, which is a pressure-retaining component, has a 103.4 mm inside diameter ${\times}$ 4.19 mm wall thickness, and is 6.36 m long, made of a zirconium alloy (Zr-2.5 wt% Nb). This provides support for the fuel while transporting the $D_2O$ heat-transfer fluid. The simple tubular geometry invites highly automated inspection, and good approach for all inspection. Similar to all nuclear heat-transfer pressure boundaries, the PHWR pressure tube requires a rigorous, periodic inspection to assess the reactor integrity in accordance with the Korea Nuclear Safety Committee law. Volumetric-based nondestructive evaluation (NDE) techniques utilizing ultrasonic and eddy current testing have been adopted for use in the periodic inspection of the fuel channel. The eddy current testing, as a supplemental NDE method to ultrasonic testing, is used to confirm the flaws primarily detected through ultrasonic testing, however, eddy current testing offers a significant advantage in that its ability to detect surface flaws is superior to that of ultrasonic testing. In this paper, effectiveness of flaw detection and the depth sizing capability by eddy current testing for the inside surface of a pressure tube, will be introduced. As a result of this examination, the ET technique is found to be useful only as a detection technique for defects because it can detect fine defects on the surface with high resolution. However, the ET technique is not recommended for use as a depth sizing method because it has a large degree of error for depth sizing.