• Korean Journal of Materials Research
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• v.14 no.10
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• pp.713-718
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• 2004

To study the dynamic strain aging behavior of Zr-0.4Sn-1.5Nb-0.2Fe sample tube for nuclear fuel cladding in the range of pressurized water reactor (PWR) operation temperature, the tensile tests of the tube specimens, which had been finally heat-treated at $470^{\circ}C\;and\;510^{\circ}C$, had been carried out with the strain rate $1.67{\times}10^{-2}/s\;and\;8.33{\times}10^{-5}/s$ at the various temperatures from room temperature to $500^{\circ}C$. It was observed that the elongation of the specimens got shortened as the temperature increased from $200^{\circ}C\;to\;340^{\circ}C$. The specimens that were finally heat-treated at $470^{\circ}C$ showed a plateau more remarkably on the plot of yield strength-temperature than those heat-treated at $510^{\circ}C$. In the range of $310\sim400^{\circ}C$, the strain rate sensitivity of the specimens finally heat-treated at $510^{\circ}C$ was $30.4\%\sim33.7\%$ lower but the work hardening exponent index of the specimens was a little higher than that without dynamic strain aging effect.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.11a
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• pp.19-26
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• 2001

The wear characteristic of Zircaloy-4 tube, which is used for a cladding of light water reactor fuel rod, is investigated experimentally. The experiment is conducted with contacting the crossed tube specimens in air as well as in water at room temperature with various combination of contact normal force and sliding distance of reciprocating motion. The contour and the volume of each wear are examined to study the effect of contact condition and environment on wear. As a result, it is found that the wear volume in the water environment is larger than that in the air for all the contact (i.e., force and sliding distance) conditions. However, the wear depth is greater in air than in water if the contact normal force and the sliding distance are larger. These are explained by the ease of detachment of wear particles from the contact surface. On the other hand, workrate model is applied with the contact shear force range measured by our wear tester. Investigated is the correlation between the workrate and the wear volume increase rate of the present experiment. The parabolic curve is found to fit well for the present wear data.

• Korean Journal of Materials Research
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• v.12 no.2
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• pp.140-145
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• 2002

This study was conducted to investigate the brazing characteristics between Zircaloy-4 nuclear fuel cladding tubes and bearing pads with filler metals of amorphous $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.5) binary alloy, in which they were produced in the ribbon form by the melt-spinning metod. The crystallization behavior, stability, hardness and micro-structure of brazed zone were examined by X-ray diffraction, differential scanning calorimetry, micro-Vickers hardness test, optical microscopy, and transmission electron microscopy. $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.4) amorphous alloys were crystallized to $\alpha$-Zr with increasing the temperature, and the rest were transformed to ZrBe$_2$at higher temperatures. On the other hand, $Zr_{1-x}Be_x$(0.4$\leq$x$\leq$0.5) amorphous alloys were crystallized to $\alpha$-Zr and ZrBe$_2$, simultaneously. The thickness of the layer brazed with amorphous alloy was increased with increasing the beryllium content due to the higher diffusion of Be. The morphology of brazed layer with PVD Be filler metal showed dendrite while that brazed with amorphous alloys appeared globular. Micro-Vickers hardness of brazed zone increased as the beryllium content of filler metal was decreased.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

• 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.

• Journal of Welding and Joining
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• v.9 no.4
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• pp.60-68
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• 1991

The weld quality of end cap welds in Pressurized Heavy Water Reactor (PHWR) Fuel is extremely important for the fuel performance in the nuclear reactor. The quality of resistance upset welds is currently evaluated mainly by the metallographic examination although it reveals only two weld cross-sections in a circumference welds. This investigation was, firstly, carried out to determine whether the ultrasonic examination would be applied to detect weld defects in the end cap welds and, secondly, to measure the mechanical strength of upset butt welds as a function of phase shift percentage. The major results obtained in this study are as follows: 1. The weld current and amount of upset shrinkage linearly increased with increasing the phase shift percentage. 2. Above the phase shift 55%, the defects in the welds were completely eliminated with increasing the phase of sound weld was over the thickness of cladding tube. 3. The ultrasonic testing well detected such defects in the end cap welds as upset external crack, upset split, corner crack and irregular weld flash comparing with the results of metallography. 4. The micro-fissure in the corner of the end cap welds was reliably detected by ultrasonic testing. 5. The mechanical strength in the welds increased with increasing phase shift percentage but the fracture did't occur in the welds above 55%.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.1-6
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• 2024

Crud is a corrosion deposit that forms in equipments and piping of nuclear reactor's primary systems. When crud circulates through the reactor's primary system coolant and adheres to the surface of the nuclear fuel cladding tube, it can lead to the Axial Offset Anomaly (AOA) phenomenon. This occurrence is known to potentially reduce the output of a nuclear power plant or to necessitate an early shutdown. Consequently, worldwide nuclear power plants have employed ultrasonic cleaning methods since 2000 to mitigate crud deposition, ensuring stable operation and economic efficiency. This paper details the system configuration of ultrasonic nuclear fuel cleaning equipment, outlining the function of each component. The objective is to contribute to the local domestic production of ultrasonic nuclear fuel cleaning equipment. Additionally, the paper introduces a method for accurately measuring the weight of removed crud, a crucial factor in assessing cleaning effectiveness and providing input data for the BOA code used in core safety evaluations. Accurate measurement of highly radioactive filters containing crud is essential, and weighing them underwater is a common practice. However, the buoyancy effect during underwater weighing may lead to an overestimation of the collected crud's weight. To address this issue, the paper proposes a formula correcting for buoyancy errors, enhancing measurement accuracy. This improved weight measurement method, accounting for buoyancy effects in water, is expected to facilitate the quantitative assessment of filter weights generated during chemical decontamination and system operations in nuclear power plants.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.179-186
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• 2009

This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

• Journal of Mechanical Science and Technology
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• v.15 no.9
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• pp.1274-1280
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• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• Journal of the Korean institute of surface engineering
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• v.35 no.5
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• pp.312-318
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• 2002

A study on the change of mechanical properties and oxidation behavior of Zircaloy-4 fuel cladding after exposing at 90$0^{\circ}C$ and $1000^{\circ}C$ for various periods of exposure time under the steam atmosphere was carried out. The growth of the $ZrO_2$ layer combined with an oxygen-rich-phase layer into the Zircaloy tube material can be described by an expression, E = 1.1√Dt + $2 $\times$ 10^{-4}$ . The tensile strength of Zircaloy tubes increased for a short period of exposure time and decreased rapidly with further exposure while the hoop strength was not decreased greatly. In the meantime, the axial and circumferential elongations of oxidized Zircaloy tubes were decreased drastically with increasing exposure time as a result of embrittlement phenomena.