• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.821-830
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• 2009

In this paper, modal testing and finite element modeling results to identify the modal parameters of a nuclear fuel rod as well as its cladding tube are discussed. A vertically standing full-size cladding tube and a fuel rod with lead pellets were used in the modal testing. As excessive flow-induced vibration causes a failure in fuel rods, such as fretting wear, the vibration level of fuel rods should be low enough to prevent failure of these components. Because vibration amplitude can be estimated based on the modal parameters, the dynamic characteristics must be determined during the design process. Therefore, finite element models are developed based on the test results. The effect of a lumped mass attached to a cladding tube model was identified during the finite element model optimization process. Unlike a cladding tube model, the density of a fuel rod with pellets cannot be determined in a straightforward manner because pellets do not move in the same phase with the cladding tube motion. The density of a fuel rod with lead pellets was determined by comparing natural frequency ratio between the cladding tube and the rod. Thus, an improved fuel rod finite element model was developed based on the updated cladding tube model and an estimated fuel rod density considering the lead pellets. It is shown that the entire pellet mass does not contribute to the fuel rod dynamics; rather, they are only partially responsible for the fuel rod dynamic behavior.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• Korean Journal of Materials Research
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• v.29 no.8
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• pp.483-490
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• 2019

Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.931-932
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• 2004

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.4
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• pp.491-501
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• 2021

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1377-1380
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• 2008

An irradiation test of a nuclear fuel rod having external and internal tube structure was planned for a performance. To establish fabrication process satisfying the requirements of irradiation test, micro-TIG welding system for fuel rods was developed, and preliminary welding experiments for optimizing process conditions of fuel rod was performed. Fuel rods with 15.9mm diameter and 0.57mm wall thickness of cladding tubes and end caps have been used and optimum conditions of endcap welding have been selected. In this experiment, the qualification test was performed by tensile tests, helium leak inspections, and metallography examinations to qualify the endcap welding procedure. The soundness of the welds quality of a dual cooled fuel rods has been confirmed by mechanical tests and microstructural examinations.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2990-3002
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• 2021

The mechanical interaction between the fuel pellet and the cladding tube of a nuclear fuel rod is a very important for safety studies as this phenomenon could lead to fuel failure and release of radioactivity. To investigate the ductility of cladding tubes used in WWER type nuclear power plants, several mandrel tests were performed in the Centre for Energy Research (EK). This modified mandrel test was used to model the mechanical interaction between the fuel pellet and the cladding using a segmented tool. The tests were conducted at room temperature and at 300 ℃ with inactive as-received and hydrogenated cladding ring samples. The results show a gradual decrease in ductility as the hydrogen content increases, the ductile-brittle transition was seen above 1500 ppm hydrogen absorbed.

• Tribology and Lubricants
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• v.24 no.4
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• pp.186-189
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• 2008

The flow induced vibration in the nuclear fuel assembly causes the fretting wear between the fuel cladding tubes and the supporting grids. The reduction in tube thickness due to the fretting wear could be related to the serious damage on nuclear fuel assembly. In this paper, the effect of the water flow on fretting wear of nuclear fuel cladding tube against supporting grid was investigated through the fretting wear tester with water spout equipment. The test results were compared with the data conducted in the stationary water. At stationary water environment the wear debris was trapped between fretting surfaces, and then the fretting wear occurred by three-body abrasion. However, in the case of water flow, the two-body abrasive wear was the dominant wear mechanism, because the wear debris was easily removed by water flow.

• Tribology and Lubricants
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• v.23 no.3
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• pp.130-133
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• 2007

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube and support. The fretting wear of tube and support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.

• Tribology and Lubricants
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• v.24 no.3
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• pp.129-132
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• 2008

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube and support. The fretting wear of tube and support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.