• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.329.1-329
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.256.1-256
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.262.2-262
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.239-239
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• 2002

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.746-751
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• 2001

Design requirements for the nuclear fuel assembly grid of the pressurized water reactor are reviewed from the mechanical/structural point of view. And mechanical/structural tests and numerical analyses on the various spacer grid candidates that has been uniquely designed by KAERI are carried out to find out their mechanical/structural performance. As a result, the results from the numerical analyses are good agreements with test results.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.2
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• pp.201-206
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• 2011

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a pressurized water reactor (PWR), and it affects the performance of the fuel assembly. The primary design requirement is that the mechanical integrity of the fuel rod should be maintained by the spacer grid assembly during the operation of the reactor. It was known that fretting damage to the fuel rod can be reduced by adjusting the relative moving displacement between the fuel rod and its support. In this study, we used the finite element method to evaluate the characteristics of a sliding support designed to reduce fretting damage of fuel rods.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.355-360
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• 2001

Design requirements for the nuclear fuel assembly grid of the pressurized light water reactor(PLWR) are scrutinized from the mechanical/structural point of view. As a result of the scrunity, mechanical/structural test facilities on the spacer grid of the PLWR Fuel are set up in KAERI to find out their mechanical/structural performance.

• 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.28 no.1
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• pp.17-26
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• 1996

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.