• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.383.2-383
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.185.2-185
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.186.1-186
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.330.2-330
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.119-119
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.132.1-132
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.230.2-230
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.13-14
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• 2005

KAERI is performing a project on out-pile test technology development for a full scale PWR fuel assembly. We have introduced the hydraulic test facility, a test assembly, test parameters, test methods, and a data acquisition system. The start up test will be in the middle of March 2005 and the main test will be accomplished by the end of 2006. The established test facility and measuring technique will contribute to the satisfaction of domestic needs for the design verification to improve the reliability of a PWR plant operation.

• Journal of Welding and Joining
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• v.24 no.5
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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.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.330-338
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• 2023

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.