• 한국소음진동공학회논문집
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• 제16권2호
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• pp.163-168
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• 2006

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle $(5\times5)$ which is called partial fuel assembly is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid to obtain the Flow-Induced Vibration (FIV) characteristics of the scaled fuel assembly over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the assembly prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the scaled test assembly. For the damping factor of the partial fuel assembly and the grid cage at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the partial fuel assembly is about $0.7\%$ with reasonable error of $2\%$ for the previous results. Nonlinear behavior of the partial fuel assembly might be stem mainly from the rod-grid support configuration.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2006년도 춘계학술대회논문집
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• pp.246-249
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• 2006

Finite element vibration analysis of the trial 5x5 partial fuel assembly in the still water was performed using the substructure method. ANSYS software was used as a finite element modeling and modal analysis tool. The calculated natural frequencies of the partial fuel assembly were more consistent with the experimental results for the identical test model compared to the much larger solid model. This modeling technique can be utilized for the fuel assembly dynamic behavior analysis under normal operation, seismic and loss-of-coolant-accident analysis.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3943-3948
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• 2022

Single-photon emission computed tomography is one of the reliable pin-by-pin verification techniques for spent-fuel assemblies. One of the challenges with this technique is to increase the total fuel assembly verification speed while maintaining high verification accuracy. The aim of the present study, therefore, was to develop an artificial intelligence (AI) algorithm-based tomographic image analysis technique for partial-defect verification of fuel assemblies. With the Monte Carlo (MC) simulation technique, a tomographic image dataset consisting of 511 fuel-rod patterns of a 3 × 3 fuel assembly was generated, and with these images, the VGG16, GoogLeNet, and ResNet models were trained. According to an evaluation of these models for different training dataset sizes, the ResNet model showed 100% pattern estimation accuracy. And, based on the different tomographic image qualities, all of the models showed almost 100% pattern estimation accuracy, even for low-quality images with unrecognizable fuel patterns. This study verified that an AI model can be effectively employed for accurate and fast partial-defect verification of fuel assemblies.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.211.1-211
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.299-299
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.839-840
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.300.1-300.1
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.459-460
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• 2005

• Nuclear Engineering and Technology
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• 제21권3호
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• pp.193-204
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• 1989

본 논문에서는 정보 이론의 maximum entropy Principle을 이용하여 중성자속 분포를 재생하는 새로운 방법을 시도하였다. 어떤 대상에 대한 부분적인 정보가 있을 때, 이 정보의 한도 내에서 entropy를 최대화시키는 확률 분포는 가장 객관적인 것이 된다. Nodal method계산결과인 평균 중성자속과 current의 값을 prior information으로 삼고, 핵 연료 집합체의 경계에서의 중성자속 분포를 확률의 형태로 변환해서 확률로써 다룬다. Prior information의 한도 내에서 entropy를 최대화시키는 경계에서의 확률 분포를 구하면 핵연료 집합체의 경계에서의 중성자속 분포가 구해지는데, 이것을 경계조건으로 heterogeneous assembly calculation을 행하여 세부적인 중성자속 분포를 구한다. 이 새로운 방법을 몇 개의 benchmark problem assembly에 응용해 본 결과, 노심의 안쪽 부분에서는 이 방법이 form function method에 의한 것과 비슷한 정확도를 보였고 바깥 부분에서는 다소 큰 오차를 보였다. 본 논문에서는 surface-averaged neutron current를 prior in-formation에 포함시키지 못했는데, 이것을 포함시키면 결과가 훨씬 개선 될 것으로 보인다.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.93-101
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• 2001

The spacer grid is one of the main structural components in the fuel assembly, which supports the fuel rods, guides cooling water, and protects the system from an external impact load, such as earthquakes. Therefore, the mechanical and structural properties of the spacer grids must be extensively examined while designing them. In this paper, a numerical method for predicting the buckling strength of spacer grids is presented. Numerical analyses on the buckling behavior of the spacer grids are performed for a various array of sizes of the grids considering that the spacer grid is an assembled structure with thin-walled plates and imposing proper boundary conditions by nonlinear dynamic finite element method using ABAQUS/Explicit. Buckling tests on several numbers of specimens of the spacer grid were also carried out in order to compare the results between the test and the simulation result. The drop test is accomplished by dropping a carriage on the specimen at a pre-determined position. From this test, the specimens are buckled only at the uppermost and the lowermost layer among the multi-cells, which is similar to the local buckling at the weakest point of the grid structure. The simulated results also similarly predicted the local buckling phenomena and were found to give good correspondence with the experimental values for the thin-walled grid structures.