• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.463-464
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.417-418
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.342-343
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• 2005

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.369-377
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• 2003

Since Zr-2.5Nb pressure tubes have a high risk for the formation of blisters during their operation in pressurized heavy water reactors, there has been a strong incentive to develop a method for the non-destructive detection of blisters grown on the tube surfaces. However, because there is little mismatch in acoustic impedance between the hydride blisters and zirconium matrix, it is not easy to distinguish the boundary between the blister and zirconium matrix with conventional ultrasonic methods. This study has focused on the development of a special ultrasonic method, so called ultrasonic velocity ratio method for a reliable detection of blisters formed on Zr-2.5Nb pressure tubes. Hydride blisters were grown on the outer surface of the Zr-2.5Nb pressure tube using a cold finger attached to a steady state thermal diffusion equipment. To maximize a difference in the ultrasonic velocity in hydride blisters and the zirconium matrix, the ultrasonic velocity ratio of longitudinal wave to shear wave, $V_L/V_S$, has been determined based on the flight time of the longitudinal echo and reflected shear echo from the outer surface of the tubes. The feasibility of the ultrasonic velocity ratio method is confirmed by comparing the contour plots reproduced by this method with those of the blisters grown on the Zr-2.5Nb pressure tubes.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.746-755
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• 2015

To investigate the effect of hydrogen and oxygen contents on hydride reorientations during cool-down processes, zirconium-niobium cladding tube specimens were hydrogen-charged before some specimens were oxidized, resulting in 250 ppm and 500 ppm hydrogen-charged specimens containing no oxide and an oxide thickness of $0.38{\mu}m$ at each surface. The nonoxidized and oxidized hydrogen-charged specimens were heated up to $400^{\circ}C$ and then cooled down to room temperature at cooling rates of $0.3^{\circ}C/min$ and $8.0^{\circ}C/min$ under a tensile hoop stress of 150 MPa. The lower hydrogen contents and the slower cooling rate generated a larger fraction of radial hydrides, a longer radial hydride length, and a lower ultimate tensile strength and plastic elongation. In addition, the oxidized specimens generated a smaller fraction of radial hydrides and a lower ultimate tensile strength and plastic elongation than the nonoxidized specimens. This may be due to: a solubility difference between room temperature and $400^{\circ}C$; an oxygen-induced increase in hydrogen solubility and radial hydride nucleation energy; high temperature residence time during the cool-down; or undissolved circumferential hydrides at $400^{\circ}C$.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1579-1587
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• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.421-422
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• 2005

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.751-757
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• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• Korean Journal of Materials Research
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• v.15 no.2
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• pp.126-133
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• 2005

Mechanical properties of zirconium cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) at high burnup situation as an out-reactor test. Zircaloy-4 cladding was hydrided up to 1000 ppm as well as oxidized up to $100\;{\mu}m$ to simulate high-burnup situation. After simulated high-burnup treatment, ring tension test was carried out from 0.01 to 1/sec to correlate with actual RIA event. The results showed that ductility and circumferential toughness decreased with the hydrogen content and oxide thickness. Hydride generated inside cladding acted as brittle failure. Oxygen influenced cladding tube by the reduction of load bearing area, oxygen embrittlement, and thermal aging. Correlation between in-reactor RIA parameter like fuel enthalpy and out-reactor toughness was performed and showed a reasonable result.