• Tribology and Lubricants
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• v.15 no.1
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• pp.83-89
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• 1999

The fretting wear behaviour of Zircaloy-4 tube used as the fuel rod cladding in PWR nuclear power plants has been investigated at the different test environment, in light water and in air as a function of slip amplitude, normal load, test duration and frequency. Zircaloy-4 tubes were used for both of oscillating and stationary specimens. A fretting wear tester was designed to be suitable for this fretting test. The wear volume and specific wear rate of Zircaloy-4 tube in water was greater than those in air under various slip amplitude. Delaminates and surface cracks were observed at low slip amplitude and high load of fretting test in water, but the traces of adhesion and plowing were observed at and above 200 Um. The water accelerates the wear of Zircaloy-4 tube at lower slip amplitude in fretting.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.927-931
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• 2021

In this work, an attempt is made to improve the electrochemical corrosion resistance of Zircaloy-4 by helium ions implantation. For this purpose, the Zircaloy-4 was implanted with 300 keV helium ions of fluences 1 × 10¹³, 1 × 10¹⁵, and 1 × 10¹⁶ ions-cm^-2 by using Pelletron Accelerator. Electrochemical tests of pristine and ion-implanted samples were performed in NaCl solution and their potentiodynamic polarization curves were obtained. The results showed enhancement of the corrosion resistance of Zircaloy-4 after helium ions implantation. The corrosion rate and current density of the material were significantly reduced by the helium implantation. The decrease in corrosion parameters was attributed to helium ions diffusion inside Zircaloy-4 that reduced the electrons flow from the samples.

• Journal of Welding and Joining
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• v.11 no.4
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• pp.70-78
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• 1993

An attempt was made to investigate the effect of brazing time on microstructure, microhardness, and corrosion of Zircaloy -4as well as the beryllium diffusion into its sheet. The sheets were coated with beryllium and brazed at $1020^{\circ}C$ for 20-40 minutes in $2{\times}10^{-5}$ torr vacuum atmosphere. 1. Microstructurally the brazed zone was largely divided into three regions: a region of continuous or partially formed of eutectic liquid films along grain boundaries; a region of precipitation in both grains and grain boundaries; a region of elongated wide structure of .alpha.-laths, which was not affected by beryllium. 2. Due to the precipitates, the beryllium-migrated region was hardened and the width of the hardened region increased with increasing brazing time. 3. Beryllium brazed Zircaloy -4 sheets showed a higher corrosion rate than those of as-received and heat-treated at a brazing temperature. 4. Diffusion coefficient of beryllium into Zircaloy -4 at $1020^{\circ}C$ for 30 minutes was $7.67{\times}10^{-7}cm^2/sec.$ It seemed that Be penetrated Zircaloy -4 by forming eutectic liquid films along grain boundaries in the proximity of Be/Zr interface and it, thereafter, diffused into Zircaloy mainly by interstitial solid solution.

• Proceedings of the KWS Conference
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• 2000.04a
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• pp.139-143
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• 2000

Laser beam weldability of Zircaloy-4 was investigated using a pulsed Nd:YAG laser of 550W average power. Mechanical properties and microstructure of laser butt welded Zircaloy-4 test specimens were examined. The influence of laser generated during laser welding was analyzed and optimum laser welding parameters were investigated.

• Journal of Welding and Joining
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• v.14 no.6
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• pp.81-92
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• 1996

This study is to compare the weldability of Zircaloy-4 end cap of nuclear fuel pin using by GTA and Nd :YAG LB. The welding parameters which affect bead width and penetration depth have been investigated. The effect of joint geometry of end cap for GTAW and LBW has been studied and optimum conditions of Zircaloy-4 endcap welding have been found. Microstructures and microhardness of GTA and LB welded zones have been also compared.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.

• Journal of the Korean institute of surface engineering
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• v.35 no.6
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• pp.363-370
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• 2002

Empirical equations for the oxide thickness and the weight gain of Zircaloy-4 cladding during the oxidation in high temperature, high pressure steam have been developed. Firstly, the empirical equations for oxide thickness in 1 atm steam in 700~100$0^{\circ}C$ were made, then, the enhancement factor for the steam pressure effects on Zircaloy-4 cladding oxidation in high temperature steam was added. Based on the analysis of the weight fraction of dissolved oxygen in metal layer, empirical equations for the weight gain of Zircaloy-4 in high pressure, high temperature steam were developed. We compare the developed empirical equations with the Baker-Just correlation. The Baker-Just correlation can give a non-conservative estimation of oxidation of Zircaloy-4, depending on the steam pressure. These developed empirical equations can be used for the correct estimation of oxidation of Zircaloy-4 during accident analysis.

• Journal of the Korean institute of surface engineering
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• v.23 no.1
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• pp.44-51
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• 1990

Effects of temperature on the tensile properties of annealed and hydirded Zircaloy-4 plate in which hydrides are precitated paralled to the rolling direction were investigated. The main results obtained are as follows : 1) In annealed Zircaloy-4, yield point phenomenon was found in the temperature range of $200-550^{\circ}C$, while in hydrided alloy the phenomenon was found in the range of $200-400^{\circ}C$. 2) The dynamic strain aging behavior was occurd in the temperature interval of 400-$550^{\circ}C$in both annealed and hydrided Zircaloy-4. 3) The nearly values of yield strength, tensile stength and elongation are obtained in both annealed and hydried Zircaloy-4. From this result, we are led to conclude that the hydrides which are preiptated parallel to the circumferenial direction of nucler fuel are not so harmful for tensile properties of the clad.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.173-179
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• 1998

The corrosion behavior of KOFA cladding which is a standard Zircaloy-4 manufactured by Westinghouse Specialty Metal Plant according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification was analyzed using the oxide measurement data of KOFA fuel irradiated in Kori-2 nuclear power plant. Analysis of the measured KOFA cladding oxidation showed that oxidation of KOFA cladding was lower than the design prediction based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Although the measured fuel rods have relatively low burnup and oxidation and the amount of the measured data are small, analysis of manufacturing and in-reactor operation conditions of KOFA cladding indicates that the differences in the manufacturing processes and chemical composition of the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 and KOFA cladding may have somewhat contributed to lower corrosion of KOFA cladding than the expected.

• Korean Journal of Materials Research
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• v.8 no.11
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• pp.1031-1037
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• 1998

To investigate the effect of cooling rate and annealing temperature on the corrosion of Zircaloy-4 and Zr-2. 5Nb alloys, autoclave corrosion tests were performed at $500^{\circ}C$ for the specimens prepared by various heat treatments. The specimens were heat-treated at $1050^{\circ}C$ for 30 minutes and cooled by ice-brine quenching, water quenching, oil quenching, air cooling, and furnace cooling. To investigate the effect of annealing temperature, the specimens were annealed at $\alpha$, ($\alpha$+$\beta$)-, and $\beta$-temperatures. It was observed from the $500^{\circ}C$ corrosion test that nodular corrosion occurred on the Zircaloy-4 alloy but did not occur on the Zr-2.5Nb alloy. The corrosion resistance of Zircaloy-4 increased with increasing the cooling rate. On the other hand, the corrosion resistance of Zr-2.5Nb decreased with increasing the cooling rate and the annealing temperature. It is suggested that corrosion resistance of Zircaloy-4 would be controlled by the distribution of Fe and Cr element in the matrix and precipitates, while that of Zr-2.5Nb alloy the niobium concentration and $\beta_{-Nb}$ phase.