• Journal of Surface Science and Engineering
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• v.47 no.1
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• pp.7-12
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• 2014

The characteristics of oxidation for the Zry-4 was measured in the $800^{\circ}C$ and high steam pressure (50 bar, 75 bar, 100 bar) conditions, using an apparatus for high pressure steam oxidation. The effect of accelerated oxidation by high-pressure steam was increased more than 60% in hydrogen-charged cladding than normal cladding. This difference between hydrogen charged claddings and normal claddings tends to be larger as the higher pressure. The accelerated oxidation effect of hydrogen charging cladding is regarded as the hydrogen on the metal layer affects the formation of the protective oxide layer. The creation of the sound monoclinic phase in Zry-4 oxidation influences reinforcement of corrosion-resistance of the oxide layer. The oxidation is estimated to be accelerated due to the creation of equiaxial type oxide film with lower corrosion resistance than that of columnar type oxide film. When tetragonal oxide film transformed into the monoclinic oxide film, surface energy of the new monoclinic phase reduced by hydrogen in the metal layer.

• Corrosion Science and Technology
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• v.18 no.5
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• pp.175-181
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• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.680-686
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• 1995

It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors ill the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.844-847
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• 2002

This paper is to investigate the wear behavior of submerged-arc clad materials by the wear test with a ball-on-disk type wear testing machine in air. The specimens were clad with Stoody105 alloy wire on a carbon steel (SM45C) substrate by submerged-arc cladding process under different welding parameters. The wear behavior of the cladding through ball-en-disk test has been studied under the wear load from 5N to 16N and sliding speed from 8cm/s to 35cm/s. The weight of the specimen loss was measured. Scanning electron micrographs of the worn surface show a layer of oxide film formed on the worn surface. Oxidation wear mechanism controls the wear process. The spalling of the oxide is caused by the repeated rubbing fatigue mechanism.

• Journal of the Korean Society for Precision Engineering
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• v.20 no.1
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• pp.179-186
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• 2003

This paper is to investigate the wear behavior of submerged-arc claddings by the wear test with a ball-on-disk type wear testing machine in air. The specimens were clad with Stoody105 alloy wire on a medium carbon steel (SM45C) substrate by submerged-arc cladding process under different welding parameters. The wear behavior of the cladding through ball-on-disk test has been studied under the wear load from 5 to 16 N and the sliding speed from 8 to 35 cm/s. The weight loss of the specimen was measured. Scanning electron micrographs of the worn surface show a layer of oxide film formed on the worn surface. Oxidation wear mechanism controls the wear process. The spatting of the oxide is caused by the repeated rubbing fatigue mechanism.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.163-170
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• 2009

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

• Korean Journal of Materials Research
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• v.14 no.7
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• pp.462-469
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• 2004

The effect of Cu addition on the corrosion characteristics of Zr alloys that developed for nuclear fuel cladding in KAERI (Korea Atomic Energy Research Institute) was evaluated. The alloys having different element of Nb, Sn, Fe, Cr and Cu were manufactured and the corrosion tests of the alloys were performed in static autoclave at $360^{\circ}C$, distilled water condition. The alloys were also examined for their microstructures using the optical microscope and the TEM equipped with EDS and the oxide property was characterized by using X-ray diffraction. From the result of corrosion test more than 450 days, the corrosion rate of the Zr-based alloys was changed with alloying element such as Nb, Sn, Fe, Cr and especially affected by Cu addition. The corrosion resistance was increased with increasing the Cu content and the tetragonal $ZrO_2$ layer was more stabilized on the Cu-containing alloys.

• Journal of the Institute of Electronics Engineers of Korea SD
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• v.41 no.9
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• pp.31-37
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• 2004

We have proposed a design methodology for large active-area single-mode VCSELS, which have higher reliability and output power, and are well-suited for high-speed operation. The key idea underlying the design methodology is to reduce the effective index difference between active and cladding regions by controlling the thickness and position of the oxide layer. The idea is confirmed by the self-consistent effective index method. By placing the oxide layer position properly, we can increase the radius of the oxide aperture for single-mode operation by 3 times.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.526-535
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• 1994

For the evaluation of the corrosion behavior of the aluminum cladding in the KMRR(Korea Multipurpose Research Reactor) fuel, a modified Griess correlation was derived by introducing a heat flux factor derived from the comparison of the measured in-reactor corrosion data with the prediction of the Griess correlation. As a design criterion on the corrosion to maintain the KMRR fuel integrity, prevention of the oxide spallation was conservatively selected, which is conservatively assumed to occur when the temperature difference across the oxide layer exceeds 114$^{\circ}C$. A bounding power history of the KMRR fuel was determined by examining all the power histories of the KMRR fuel from cycle 1 to equilibrium cycle, and used to predict the maximum possible corrosion. Results of the corrosion prediction of the KMRR fuel with the bounding power history showed that the maximum local thickness of the oxide layer would be below 50$\mu$m and the design criterion on the oxide spallation would be satisfied with a factor of two margin. Therefore, it can be said that corrosion of the cladding will not impair the integrity of the KMRR fuel. Nevertheless, the applicability of the modified Griess correlation to the KMRR needs to be further verified through the KMRR fuel corrosion surveillance.