• Nuclear Engineering and Technology
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• v.18 no.4
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• pp.267-272
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• 1986

To obtain various necessary data for the manufacturing and the use of the nuclear fuel cladding tube, the effects of deformation and heat treatment on Properties of Zircalof-4 material have been studied. The hardness is increased rapidly at a low degree of cold work and increased rapidly at cold work above 10%. Recrystallization has been completed at 64$0^{\circ}C$, 59$0^{\circ}C$, and 555$^{\circ}C$ in 30%, 60% and 80% cold worked specimen, respectively. The transformation of microstructure with increasing cooling rate after $\beta$-annealing is as follows; coarse Widmanstatten ($\alpha$) longrightarrow fine parallel plate ($\alpha$) longrightarrow martensite ($\alpha$$^{'}$). At the same time, hardness increased with increasing cooling rate. rate.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.249-258
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• 2010

Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at $400^{\circ}C$ and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.381-393
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• 1993

This paper presents the experimental results on flooding of countercurrent flow in vertical multirod channels, which consists of falling water film and upward air flow. In particular, the effects of spacer grids, with and without mixing vane, and of blockage in the multirod bundle on the behaviour of flooding were investigated. The 5$\times$5 zircaloy tube bundle was used for the test section. The comparison of previous analytical models and empirical correlations with present data on flooding showed that the existing models and correlations predict much higher flooding curves. The spacer grid causes the lower flooding air flow rate to compare with the bare rod bundle. However, the mixing spacer grids need a higher flooding air flow rate for a constant liquid flow rate than the spacer grids without mixing vanes. The bundle containing blockages has the highest flooding air flow rate among the bundles with spacer grids and blockages. Empirical flooding correlations for the three types of test section have been made.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.