• Nuclear Engineering and Technology
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• 제48권1호
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• pp.33-42
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• 2016

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the "volume-preserving" streamlined heterogeneous spacer grids, but the "banded" dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.

• Nuclear Engineering and Technology
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• 제33권3호
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• pp.283-297
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• 2001

The CHF in PWR rod bundles is usually predicted by the local flow correlation approach based on subchannel analysis while difficulty exists due to the existence of spacer grids especially with mixing vanes. In order to evaluate the effect of spacer grids on CHF, the experimental rod bundle data with various types of spacer grids were analyzed using the subchannel code, COBRA-IV-i. For the Plain grid data, a CHF correlation was described as a function of local flow conditions and heated length, and then the residuals of the CHF in mixing vaned grids predicted by the correlation were examined in various kinds of grids. In order to compensate for the residual, three parameters, distances between grids and from the last grids to the CHF site, and equivalent hydraulic diameter were introduced into a grid parameter function representing the remaining effect of spacer grids predicted most of the CHF data points in plaing grids within $\pm$20 percent error band. Good agreement with the CHF data was also shown when the grid parameter function for mixing vaned grids of a specific design was used to compensate for the residuals of the CHF data predicted by the correlation.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2006년도 춘계 학술대회 개요집
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• pp.39-41
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• 2006

The weld quality of spacer grids in Pressurized Water Reactors(PWR) fuel is extremely important for the fuel assembly performance in the nuclear renter. The spacer grid welds are currently evaluated mainly by the metallographic examination although it reveals only cross-points which are welded by the laser beam. This experiment is also to compare the weldability of Zircaloy-4 spacer grids using by the GTA and laser beam. The effect of node geometries of spacer grids for GTAW and LBW has been studied and optimum conditions of spacer grid welding have been found. Microstructures and micro-hardness of GTA and laser beam welded zones have been also compared.

• Journal of Welding and Joining
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• 제24권5호
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• pp.22-28
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• 2006

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 for Pressurized Water Reactors (PWRs). The weld quality of spacer grids in PWRs fuel is extremely important for the fuel assembly performance in the nuclear renter. The spacer grid welds are currently evaluated mainly by the metallographic examination although it reveals only cross-points which are welded by the laser beam. This experiment is also to compare the weldability of Zircaloy-4 spacer grids using by the GTA and LB. The effect of node geometries of spacer grids for the GTAW and LBW has been studied and optimum conditions of spacer grid welding have been found. Microstructures and micro-hardness of the GTA and LB welded zones have been also compared.

• 한국전산구조공학회논문집
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• 제18권4호통권70호
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• pp.405-416
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• 2005

본 연구에서는 경수로용 핵연료집합체의 전체지지격자(Full Size Grid)와 부분지지격자(Small Size Grid)에 대한 정적 좌굴강도 실험과 전체 지지격자와 부분지지격자를 구성하는 지지격자판(Grid Strap)에 대한 정적 좌굴해석을 수행하여 지지격자의 좌굴특성을 분석하였으며, 분석결과를 이용하여 전체지지격자와 부분지지격자에 대한 좌굴하중값의 예측 가능성을 평가하였다. 좌굴강도 실험은 웨스팅하우스형 연료의 $17{\times}17$셀을 갖는 전체지지격자와 $1{\times}1,\;1{\times}2,\;1{\times}3,\;1{\times}4,\;1{\times}5,\;1{\times}17\;,2{\times}1,\;2{\times}2,\;2{\times}3,\;2{\times}9,\;2{\times}17,\;3{\times}17$ 등의 셀을 갖는 부분지지격자에 대하여 수행하였으며, 실험결과를 이용하여 지지격자의 좌굴강도와 지지격자의 행(rows)과 열(columns) 사이의 관계식을 제시하였다. 좌굴강도 해석은 범용 유한요소해석코드인 ANSYS를 이용하여 수행하였으며, 해석결과를 이용하여 지지격자의 좌굴특성을 분석하고 실험결과와 비교평가 하였다.

• 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.

• Nuclear Engineering and Technology
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• 제30권4호
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• pp.328-341
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• 1998

This paper presents in detail the hydraulic characteristic measurements using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids and has a cylindrical configuration. Axial velocity and turbulent intensity were measured. The effects of the spacer grids on the turbulent flow were investigated using the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of the fuel bundle and the loss coefficients for the spacer grids were estimated from the measured pressure drops. The turbulent thermal mixing phenomena were discussed.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2396-2401
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• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 춘계학술대회논문집
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• pp.705-709
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• 2002

The spacer grids in nuclear fuel assembly locate and align the fuel rods with respect to each other. They provide axial and lateral restraint against an excessive rod motion mainly caused by coolant flow. It is understood that each rod Is supported by multiple spacer grid. In such a case, it is important to determine spacer grid span so as to avoid resonance between the natural frequency of the fuel rods and excitation frequency. Actually dynamic characteristics of the fuel rods can be improved by assigning adequate spacer grid locations. When a dynamic performance of the structure is to be improved, design sensitivity analysis plays an important role as like many structural redesign problems. In this work, a shape design concept, different from conventional design, was applied to the problem. According to the theory shape can be a design parameter and optimal shape design can be found. This study concentrates on eigenvalue design sensitivity of the fuel rod supported by multiple spacer grids to determine optimal spacer grids positions.

• 전기학회논문지
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• 제61권12호
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• pp.1948-1952
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• 2012

The paper presents modal testing and analysis in order to obtain the dynamic characteristics of a middle spacer grids of a nuclear fuel rod. A spacer grid is one of the important structural elements supporting nuclear fuel rods. Such a fuel rod can be oscillated by its thermal expansion, neutron irradiation and etc. due to cooling water flow under the operation of a nuclear power plant. When the fuel rod vibrates, fretting wear due to repeated friction motion between the fuel rods and spacer grids can be occurred, and so the fuel rod is damaged. In this paper, through modal analysis and testing, natural frequencies and modes of a middle spacer grid were calculated, and the following conclusions were obtained. Firstly the numerical first-seven natural frequencies for spacer grids of a fuel rod having complicated structures have a small difference within 3.8% with experimental natural frequencies, and so the suitability of simulation results was verified. Secondly, experimental mode shapes for a middle spacer grid of a nuclear fuel rod were verified by obtaining lower non-diagonal terms through MAC(Modal Assurance Criteria), and were confirmed by the simulation modes.