• 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.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.05a
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• pp.82-87
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• 2003

Nuclear fuel with a big slenderness ratio is susceptible to flow-induced vibration under very severe conditions of high temperature, high flow and exposure to irradiation in nuclear reactor. The fuel assembly should, therefore, be designed to escape any resonance due to the vibration during the reactor operation, in particular, in case of the design changes. In addition, the amplitudes due to the grid vibration, the fuel rod vibration and the fuel assembly vibration should be minimized to reduce the grid-to-rod fretting wear. Fuel assembly vibration tests in air at room temperature and in water at high temperature have been performed to investigate fuel vibration behaviors. The frequency and damping during the test in air have been compared to those in water. Through the hydraulic test, the advanced assembly has been evaluated not to be susceptible to any resonance. In addition, the test data from the tests can be used to make fuel model and to evaluate grid-to-rod fretting wear.

• 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.

• 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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• 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.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3563-3579
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• 2021

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.54-60
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• 2011

A spacer grid, which supports nuclear fuel rods laterally and vertically with a friction grip, is one of the most important structural components in a PWR fuel. The form of grid strap and supporting parts such as grid spring and dimple is known to be closely related with the mechanical/structural performance of spacer grid and nuclear fuel assembly. In this study, reviewing various research results for enhancing the performance of the spacer grid, some structural design considerations and research directions on the spacer grid assembly are suggested for further study.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.75-82
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• 2011

The neutron multiplication factor in spent fuel storage racks, in which a stainless steel plate encloses a fuel assembly, was evaluated according to the variation of distance between the fuel assembly and stainless steel plate, as well as the pitch. The stainless steel plate position with the lowest multiplication factor on each pitch consistently appeared as 6mm or 9mm away from the outmost surface of the fuel assembly. Because the stainless steel plate has a thermal neutron absorption cross section, its ability to absorb neutrons can work best only if it is installed at the position where thermal neutrons can be gathered most easily. Therefore, the stainless steel plate position should not be too close or too far away from the fuel assembly, but it should be kept a pertinent distance from the fuel assembly.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1556-1568
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.