• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.895-910
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• 2013

The reflood test data from the rod bundle heat transfer (RBHT) test facility showed that the grids in the upper portion of the rod bundle could become wet well before the arrival of the quench front and that the sizes of liquid droplets downstream of a wet grid could not be predicted by the droplet breakup models for a dry grid. To investigate the water droplet generation from a wet grid spacer, a viscous linear temporal instability model of the water sheet issuing from the trailing edge of the grid with the surrounding steam up-flow is developed in this study. The Orr-Sommerfeld equations along with appropriate boundary conditions for the flow are solved using Chebyshev series expansions and the Tau-Galerkin projection method. The effects of several physical parameters on the water sheet oscillation are studied by determining the variation of the temporal growth rate with the wavenumber. It is found that a larger relative steam velocity to water velocity has a tendency to destabilize the water sheet with increased dynamic pressure. On the other hand, a larger ratio of steam boundary layer to the half water sheet thickness has a stabilizing effect on the water sheet oscillation. Droplet diameters downstream of the spacer grid predicted by the present model are found to compare reasonably well with the data obtained at the RBHT test facility as well as with other data recently reported in the literature.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.230.1-230
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• 2001

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.10
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• pp.999-1008
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• 2007

The spacer grid set is a component in the nuclear fuel assembly. The set supports the fuel rods safely. Therefore, the spacer grid set should have sufficient strength for the external impact forces such as earthquake. The fretting wear occurs between the spring of the fuel rod and the spacer grid due to flow-induced vibration. Conceptual design of the spacer grid set is performed based on the Independence Axiom of axiomatic design. Two functional requirements are defined for the impact load and the fretting wear, and corresponding design parameters are selected. The overall flow of design is defined according to the application of axiomatic design. Design for the impact load is carried out by using nonlinear dynamic analysis to determine the length of the dimple. Topology optimization is carried out to determine a new configuration of the spring. The fretting wear is reduced by shape optimization using the homology theory. The deformation of a structure is called homologous if a given geometrical relationship holds before, during, and after the deformation. In the design to reduce the fretting wear, the deformed shape of the spring should be the same as that of the fuel rod. This condition is transformed to a function and considered as a constraint in the shape optimization process. The fretting wear is expected to be reduced due to the homology constraint. The objective function is minimizing the maximum stress to allow a slight plastic deformation. Shape optimization results are confirmed through nonlinear static analysis.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.400-405
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• 2001

Nozzle of the double-plated grid plays the role of the spring to support a fuel rod as well as to provide the coolant path in grid. The nozzle was known to be necessary to reduce the spring stiffness for supporting performance. In this study the contact analysis between the fuel rod and the nozzle type spacer grid was performed by using ABAQUS standard to propose the preferable shape in tenn of spring performance. Two small cuts at the upper and lower part of the nozzle appeared to have a minor effect in decreasing the nozzle stiffness. A long slot at the center of the nozzle was turned out not only to decrease the spring constant as desired but also to increase the elastic displacement.

• Proceedings of the KWS Conference
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• 2005.11a
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• pp.70-72
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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 Zircaloy spacer grid assembly welded by two 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.48 no.1
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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.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2007.04a
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• pp.145-150
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• 2007

The spacer grid set is a component in the nuclear fuel assembly. The set supports the fuel rods saftely. Therefore, the spacer gl1d set should have sufficient strength for the external impact forces. The fretting wear occurs between the spring of the fuel rod and the spacer grid due to tile flow-induced vibration. The conceptual design of the spacer grid set is performed based on the Independence Axiom of axiomatic design. Two functional requirements are defined and corresponding design parameters are selected. The overall flow of the design is defined according to the application of axiomatic design. The design for the impact load is carried out by using nonlinear dynamic analysis to determine the length of the dimple. Topology optimization is carried out to determine a new configuration of the spring. The fretting wear is reduced by shape optimization using the homology theory. In the design to reduce the fretting wear, the deformed shape of the spring should be the same as that of the fuel rod. This condition is transformed to a function and considered as a constraint in the shape optimization process. The fretting wear is expected to be reduced due to the homology constraint. The objective function is minimizing the maximum stress to allow a slight plastic deformation. Shape optimization results are confirmed through nonlinear static analysis because the contact area becomes wider.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.13 no.5
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• pp.384-392
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• 2003

The fuel rods in the Pressurized water reactor are continuously supported by a spring system called a spacer grid which is one of the main structural components for the fuel rod cluster(fuel assembly). The fuel rods vibrate within the reactor due to coolant flow. Since the vibration, which is called flow-induced vibration(FIV) can wear away the surface of the fuel rod, it is important to understand it's vibration characteristics. In this paper, the vibration analyses and the tests for the dummy rods supported by New Doublet(ND) spacer grids are described. A new FE model which reflects the contact area between the rod and ND spacer grid spring is developed to replace the previous one by which a good agreement could not be obtained with the vibration test. The natural frequency and mode shape calculated by both the Previous FE model and the new one are compared with those of experiment for a single-spanned rod supported by two ND spacer grids. The results of the new model showed good agreement with the experiment compared with those of previous model. In addition. the new FE model is applied to the vibration analysis for the dummy rod of 2.189 mm tall continuously supported by five ND spacer grids. It is also obtained that the analysis results of the new FE model well agreed to experiment ones as the single-spanned rod.

• Nuclear Engineering and Technology
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• v.33 no.3
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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.