• Nuclear Engineering and Technology
• /
• 제46권4호
• /
• pp.521-528
• /
• 2014

The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and $Y_2O_3$ particles of $10{\mu}m$ were selected for ODS treatment using LBS. Through the LBS method, the $Y_2O_3$ particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at $500^{\circ}C$ was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive $Y_2O_3$ particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

• Corrosion Science and Technology
• /
• 제2권6호
• /
• pp.253-259
• /
• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 1998년도 제27회 춘계학술대회
• /
• pp.55-63
• /
• 1998

The fretting wear characteristics of Zircaloy-4 tube in light water were investigated experimentally. A fretting wear tester was designed to be suitable for this fretting test. This study was focused on the effects due to the combination of normal load, slip amplitude and number of cycles as the main factors of fretting. The results of this study showed that the wear volume increased abruptly at slip amplitude above 100$\mu$m, which is defined as critical slip amplitude of Zircaloy-4 tube in light water, and that under 160$\mu$m the wear volume decreased as load increased at the same slip amplitude.

• Nuclear Engineering and Technology
• /
• 제52권2호
• /
• pp.352-359
• /
• 2020

The effects of hydrogen precipitation on the mechanical properties of Zircaloy-4 and Zirlo alloys were examined with uniaxial tensile tests at room temperature and at 400 ℃ and accompanying microstructural changes in the Zircaloy-4 and Zirlo alloy specimens were discussed. The elastic moduli of Zircaloy-4 and Zirlo alloys decreased with increasing hydrogen concentrations. Yield strengths of both materials tended to decrease gradually. The reductions of yield stress seems to be caused by the dissipation of yield point phenomena shown in stress-strain curves. Ultimate tensile strengths (UTS) of Zircaloy-4 and Zirlo slightly increased at low hydrogen contents, and then decreased when the concentrations exceeded 500 and 700 wppm, respectively. Uniform elongations were stable until 600 wppm and drops to 0% around 1400 wppm at room temperature.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2001년도 춘계학술대회논문집A
• /
• pp.256-261
• /
• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

• Journal of Welding and Joining
• /
• 제15권6호
• /
• pp.85-95
• /
• 1997

This study is to investigate the optimum conditions of Nd:YAG laser beam welding for Zircaloy-4 end cap closure by optical fiber transmission. Laser welding parameters which affect the penetration depth and bead width were experimentally examined using the various beam radius by the beam quality analyzer, joint geometries of end cap and the laser parameters which mean pulse width, repetition rate and pulse energy. Also, an optimum welding speed and the effect of assistant gas with varying the flow rate of He were investigated. We found that the laser average power for the end cap welding will be 230W and rotation speed must not exceed 8 RPM, the best position of focus using optical fiber with 600.mu.m will be 2 to 3mm below the surface of the material.

• 한국표면공학회지
• /
• 제40권4호
• /
• pp.180-184
• /
• 2007

A new zirconium alloy, low-Sn Zircaloy-4 was investigated to see the effects of high pressure steam on the oxidation at high temperatures. High pressure steam turned out to enhance the oxidation at high temperatures below $1000^{\circ}C$. The oxide layer groved to deviate from the uniform layer under high steam pressures, and usually cracks were found at the thicker parts in the oxide layer. High pressure steam seems to destabilize the tetragonal oxides near the metal layer, and the monoclinic oxides transformed from the destabilized tetragonal oxides are structurally not sound, resulting in enhanced oxidation under high pressure steam.

• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1579-1587
• /
• 2022

The effects of hydride amount (20-850 wppm), orientation (circumferential and radial), and temperature (room temperature, 100 ℃, 200 ℃) on the axial mechanical properties of Zircaloy-4 cladding were comprehensively examined. The fraction of radial hydride fraction in the cladding was quantified using PROPHET, an in-house radial hydride fraction analysis code. Uniaxial tensile tests (UTTs) were conducted at various temperatures to obtain the axial mechanical properties. Hydride orientation has a limited effect on the axial mechanical behavior of hydrided Zircaloy-4 cladding. Ultimate tensile stress (UTS) and associated uniform elongation demonstrated limited sensitivity to hydride content under UTT. Statistical uncertainty of UTS was found small, supporting the deterministic approach for the load-failure analysis of hydrided Zircaloy-4 cladding. These properties notably decrease with increasing temperature in the tested range. The dependence of yield strength on hydrogen content differed from temperature to temperature. The ductility-related parameters, such as total elongation, strain energy density (SED), and offset strain decrease with increasing hydride contents. The abrupt loss of ductility in UTT was found at ~700 wppm. Demonstrating a strong correlation between total elongation and offset strain, SED can be used as a comprehensive measure of ductility of hydrided zirconium alloy.

• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
• /
• pp.185-186
• /
• 2002

Fretting of fuel rod cladding material, Zircaloy-4 tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to be an ideal solution to fretting damage since fretting is closely related to wear. corrosion and fatigue. Therefore. in this study the fretting wear experiment was performed using TiAlN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaloy-4 as on of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAlN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and fretting wear mechanisms were brittle fracture and plastic flow at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher ship amplitude.

• Nuclear Engineering and Technology
• /
• 제52권9호
• /
• pp.2047-2053
• /
• 2020

Single-tube burst tests on hydrogenated Zircaloy-4 nuclear fuel cladding under simulated loss-of-coolant accident are conducted to evaluate the impact of hydrogen on burst parameters. The heating rate and initial pressure are varied from 5 K/s to 150 K/s and 5 bar-80 bar, respectively. The hydrogen concentration in the cladding is in the range of 0-2000 wppm. Burst stress is lower for hydrogenated cladding in α-phase. A significant loss of ductility is observed in α-phase and lower α + β-phase for hydrogenated cladding. However, the burst strain is higher for hydrogenated cladding in β-phase. There is a sigmoidal dependency of rupture area with initial stress and rupture area is larger for hydrogenated cladding. A novel burst stress correlation for hydrogenated Zircaloy-4 cladding has been proposed.