• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.529-540
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• 2014

In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to determine whether the fuel plate integrity is maintained in a drop accident. A fuel assembly drop accident is classified based on where the accident occurs, i.e., inside or outside the reactor, since each occasion results in a different impact load on the fuel assembly. An analysis procedure suitable for each drop situation is systematically established. For an accident occurring outside the reactor, the direct impact of a fuel assembly on the pool bottom is analyzed using implicit and explicit approaches. The effects of the key parameters, such as the impact velocity and structural damping ratios, are also studied. For an accident occurring inside the reactor, the falling fuel assembly may first hit the fixing bar at the upper part of the standing fuel assembly. To confirm the fuel plate integrity, a fracture of the fixing bar should be investigated, since the fixing bar plays a role in protecting the fuel plate from the external impact force. Through such an analysis, the suitability of an impact analysis procedure associated with the drop situation in the research reactor is shown.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1485-1494
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• 2023

The bow of the nuclear fuel assembly is a well-known phenomenon. One of the vital criteria during the history of nuclear fuel development has been fuel assembly's mechanical stability. Once present, the fuel assembly bow can lead to safety issues like excessive water gap and power redistribution or even incomplete rod insertion (IRI). The extensive bow can result in assembly handling and loading problems. This is why the fuel assembly's bow is one of the most often controlled geometrical factors during periodic fuel inspections for VVER when compared e.g. to on-site fuel rod gap measurements or other instrumental measurements performed on-site. Our proposed screening method uses existing video records for fuel inspection. We establish video frames normalization and aggregation for the purposes of bow measurement. The whole process is done by digital image processing algorithms which analyze rotations of video frames, extract angles whose source is the fuel set torsion, and reconstruct torsion schema. This approach provides results comparable to the commonly utilized method. We tested this new approach in real operation on 19 fuel assemblies with different campaign numbers and designs, where the average deviation from other methods was less than 2 % on average. Due to the fact, that the method has not yet been validated during full scale measurements of the fuel inspection, the preliminary results stand for that we recommend this method as a complementary part of standard bow measurement procedures to increase measurement robustness, lower time consumption and preserve or increase accuracy. After completed validation it is expected that the proposed method allows standalone fuel assembly bow measurements.

• Proceedings of the KWS Conference
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• v.43
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• pp.67-69
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• 2004

The fuel assemblies as the nuclear fuel for the pressurized water reactor(PWR) are loaded in the reactor core throughout the residence time of three to five years. The spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral load acting on the fuel assembly so as to keep the fuel assembly straight. To meet the requirement, integrity on the spacer grid welding parts should be carefully checked. In this study, welding quality of the spacer grid assembly welded by several welding companies are examined and compared.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.212-221
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• 1998

In the present study, a method for the dynamic analysis of a reactor core is developed. Peak responses for the motions induced from earthquake are obtained for a core model. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are investigated. Prediction of fuel assembly stress during an earthquake requires development of a fuel assembly stress analysis model capable of interfacing with the models and results discussed in the dynamic analysis of a reactor core. This analysis uses beam characteristics which describe the overall fuel assembly response. The stress analysis method and its application for the case of an increased seismic level are also presented.

• Journal of the Korean Society for Precision Engineering
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• v.22 no.3 s.168
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• pp.93-100
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• 2005

The top and bottom nozzles of PWR fuel assembly are connected by guide thimbles and an instrumentation tube that are connected with spacer grids. The fuel rods are inserted into the each cell of spacer grids. The loads acting on the fuel assembly are transmitted to the guide thimbles through the flow plate of top nozzle The axial loads applied to the fuel assembly are not equally distributed among the guide thimble due to the geometry of the top nozzle flow plate and spacer grid. In this study, the load concentration factors for the $17\times17$ fuel assembly were calculated. The analytical model fur the calculation of the load concentration factor of top nozzle flow plate was developed using ANSYS 5.6. The finite element analyses were performed using the model composed of top nozzle, guide thimble, and spacer grid. And, the analysis results were compared with the test results.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3353-3360
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• 2022

Using data from the design earthquake of the Korean standard nuclear power plant, seismic analyses of a fuel assembly are conducted in this study. The modal characteristics are used to develop an input deck for the seismic analysis. With a time history analysis, the responses of the fuel assembly in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses at the center location of the fuel assembly are obtained in the time domain, with these outcomes then used for a detailed structural analysis of the fuel rods in the ensuing analyses. The response spectra are also generated to determine the response characteristics in the frequency domain. The structural integrity of the fuel assembly can be ensured through this type of time history analysis considering the input excitations of various earthquakes considered in the design.

• Journal of Aerospace System Engineering
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• v.6 no.4
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• pp.18-23
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• 2012

The fuel boost pump for the aircraft was first indigenously developed in Korea. It is one of the core component for fuel subsystem and composed of motor assembly, impeller assembly, and body assembly with BLDC motor. It shall provide some amount of fuel to engine system continuously for any flight condition considering sudden altitude change and any attitude. This paper describes the procedures and the results for the design, the integration, and the performance analysis of the fuel boost pump.

• Journal of Mechanical Science and Technology
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• v.20 no.9
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• pp.1514-1524
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• 2006

This work presents the three-dimensional analysis of flow and heat transfer performed for a wire-wrapped fuel assembly of liquid metal reactor using Reynolds-averaged Wavier-Stokes analysis in conjunction with 557 model as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary conditions at inlet and outlet of the calculation domain. Three different assemblies, two 7-pin wire-spacer fuel assemblies and one bare rod bundle, apart from the pressure drop calculations for a 19-pin case, have been analyzed. Individual as well as a comparative analysis of the flow field and heat transfer have been discussed. Also, discussed is the position of hot spots observed in the wire-spacer fuel assembly. The flow field in the subchannels of a bare rod bundle and a wire-spacer fuel assembly is found to be different. A directional temperature gradient is found to exist in the subchannels of a wire-spacer fuel assembly Local Nusselt number in the subchannels of wire-spacer fuel assemblies is found to vary according to the wire-wrap position while in case of bare rod bundle, it's found to be constant.

• Proceedings of the KWS Conference
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• 2005.06a
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• pp.109-111
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• 2005

Nuclear fuel assemblies for pressurized water reactors(PWR) are loaded in the reactor core throughout the residence time of three to five years. A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, laser welding qualities of the spacer grid assembly welded by several welding companies, such as weld strength, weld penetration depth, and weld bead size, are examined and compared.