• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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• pp.639-646
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• 2000

Estimation for the vibration behavior of a nuclear fuel rod with its supporting structure called spacer grid has been made by the both of experimental and analytical methods in order to compare the supporting performance of two kinds of the spacer grids which have been newly developed. For the analytical method the fuel rod was modeled as a beam continuously supported by the springs of the spacer grid, and ABAQUS computer code was utilized. After a modal testing was performed for the fuel rod supported by five spacer grids, two results has been compared to justify and compensate the both methods. It has been found that the spring design of the spacer grid could give significant effect to natural frequency and vibration amplitude of the fuel rod.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.731-736
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• 2000

An optimal design method is adopted for a spacer grid in nuclear power plant. It is made of punched sheet metal process, functioning as springs and dimples supporting fuel rods. For stress analysis of the assembled fuel rod support, a typical cell out of the repeated pattern in the assembly is modeled using 4-node shell elements. A commercial code, ABAQUS, is used for detailed analysis of contacting phenomena with friction. For the optimization, design varibles are taken from geometric parameters representing the shape of the bent leaf spring part and mating contact region with fuel rod. Objective function is considered in relation to mechanical functions and durability. Maximum yon Mises stress is considered in relation to constrained contact stress.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.138-151
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• 2022

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The FRTMB module was integrated into the self-developed severe accident analysis code ISAA to simulate the PHEBUS FPT3 experiment. The predicted rupture time and temperature of the cladding were basically consistent with the measured values, which verified the correctness and effectiveness of the FRTMB module. The results showed that the rising of gas pressure in the fuel rod and high temperature led to cladding ballooning. Consequently, the cladding hoop strain exceeded the strain limit, and the cladding burst. The developed FRTMB module can be applied not only to rod-type fuel, but also to plate-type fuel and other types of reactor fuel rods. Moreover, the FRTMB module can improve the channel blockage model of ISAA code and make contributions to analyzing the effect of clad ballooning on transient and subsequent parts of core degradation.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2006.05a
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• pp.369-370
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• 2006

During demonstrations of a process conditioning spent nuclear fuels, it is necessary to transport and handle Spent fuel road cuts from Post Irradiation Examination facility to Slitting device in The hot cell. the spent fuel pellets which are highly radioactive materials are separated with its clad and are fed into the next conditioning process. For this, a spent fuel rod, 3.5 m long, is cut by 25 cm long which is suitable length for the decladding process. These rod-cuts are packed into the capsule and are moved to the ACPF(Advanced spent nuclear fuel Conditioning Process Facility). In the ACPF, Once the capsule is unloaded in the ACPF, Capsule is taken out one-by-one and installed on the decladding device. In these processes, the crushed spent fuel pellet can be scattered inside the facilities and thus it contaminate the hot cell. In this paper, we developed the specially designed transportation and storage device for spent nuclear fuel capsules.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.268-269
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• 2005

• Journal of the Korean Society for Precision Engineering
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• v.30 no.1
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• pp.113-119
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• 2013

To develop a new nuclear fuel, it needs to make a test fuel rod and carry out burn-up test in the test loop of a research reactor to check the irradiation characteristics of the nuclear fuel. At that time, several sensors such as thermocouples, LVDTs and SPNDs are needed to be attached in and out of the fuel rod and connect them with instrumentation cables. Then, the instrumentation cables deliver the signals measured by the sensors to the measuring device located outside of the reactor pool. In particular, to install a thermocouple in a fuel rod, it needs to drill off holes on the alumina blocks and sintered $UO_2$ pellets. However, because the hardness of a sintered $UO_2$ pellet is 700 Hv (or HRC 61) and that of an alumina block is 1480 Hv, a special drilling machine which adapts a diamond coated drill bit had developed. In this study, several case experiments have been carried out to find an optimal drilling condition of the drilling machine. And, using the optimal drilling condition, minimum numbers of the holes that a drill bit can drill off are verified.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.737-746
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• 2007

A spacer grid, which is one of the most important structural components in a PWR fuel assembly, supports its fuel rods laterally and vertically. Based on in-house design experience, scrutiny of the design features of advanced nuclear fuels and the patents of other spacer grids, KAERI has devised its own spacer grid shapes and acquired patents. In this study, a performance evaluation of KAERI's spacer grid shapes was carried out from mechanical/structural and thermohydraulic view points. A comparative performance evaluation of commercial spacer grid shapes was also carried out. The comparisons addressed the spring characteristics, fuel rod vibration characteristics, fretting wear resistance, impact strength characteristics, CHF enhancement, and the pressure drop level of the spacer grid shapes. The results show that the performances of KAERI's spacer grid shapes are as good as or better than those of the commercial spacer grid shapes.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.17-26
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• 1996

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1148-1153
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• 2018

The Leaking Fuel Experiment test facility was designed to simulate the activity release from spent leaking fuel rods under steady state and transient conditions in the spent fuel pool. The experimental rig included an electrically heated fuel rod with different defects and a cooling system. The fission product transport was simulated by potassium-chloride. The conductivity changes of the water in the cooling system were measured to provide information about the amount of released solution. Defects of different sizes and positions were applied, together with a wide range of rod powers to simulate decay heat. The produced data can be used for predicting the activity release from leaking fuel under storage conditions and for the interpretation of fuel examination procedures.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.365-365
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• 2001