• 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 Korean Society of Tribologists and Lubrication Engineers Conference
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• 1999.06a
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• pp.310-316
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• 1999

The fretting wear characteristics of the contact between Zircaloy-4 tube and Inconel 600 tube have investigated. Zircaloy-4 is used for fuel rod in nuclear reactor and Inconel 600 is used for tube In steam generator of nuclear power plant. A fretting wear tester was designed to be suitable for this fretting test. In this study, the number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. This study shows that the wear scar length of Zircaloy-4 and Inconel 600 increases as number of cycles, normal load and slip amplitude increase and the wear scar length of Zircaloy-4 is more longer than that of Inconel 600 due to the surface hardness.

• 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 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.

• 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.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.22-33
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• 1987

The confirmed integrity of nuclear fuel cladding materials is an important object during steady state and transient operations at nuclear power plant. In this context, the clad material yielding behavior is especially important because of pellet-clad gap expansion. During the steep power excursion, the in-pile irradiation behavior differences between uranium-dioxide fuel pellet and zircaloy clad induce the contact pressure between them. If this pressure reaches the zircaloy clad yield pressure, the zircaloy clad will be plastically deformed. After the reactor power resumed to normal state, this plastic permanent expansion of clad tube give rise to the pellet-clad gap expansion. In this paper, the simple mandrel expansion test method which utilizes thermal expansion difference between copper mandrel and zircaloy tube was adopted to simulate this phenomenon. That is, copper mandrel which has approximately three times of thermal expansion coefficient of zircaloy-4 (PWR fuel cladding material) were used in this experiment at the temperature range from 400C to 700C. The measured plastic expansion of zircaloy outer radius and derived mathematical relations give the yield pressure, yield stress of zircaloy-4 clad at the various clad wall temperatures, the activation energy of zircaloy tube yielding, and pellet-clad gap expansion. The obtained results are in good agreement with previous experimental results. The mathematical analysis and simple test method prove to be a reliable and simple technique to assess the yielding behavior and gap expansion measurement between zircaloy-4 tube and uranium-dioxide fuel pellet under biaxial stress conditions.

• 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.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.218-222
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• 2018

An oxide-dispersion-strengthened (ODS) layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide ($Y_2O_3$)-coated Zircaloy-4 tube to induce the penetration of $Y_2O_3$ particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at $380^{\circ}C$, and from 385 to 470 MPa at $500^{\circ}C$. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to $830^{\circ}C$ at a heating rate of $5^{\circ}C/s$ and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties.