• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.91-99
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• 2017

Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radio-active materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B&PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.274-277
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• 2004

The most important mode of subcritical crack growth is primary water stress corrosion crack, which was the reported mechanism from the root cause analysis of the crack in the bimetallic welds. Stress corrosion crack growth evaluations was carried out for several flaw shapes of both axial and circumferential flaws, using the steady-state stresses including residual stresses. This evaluation considered the possibility of additional flaws in the primary loops of nuclear power plant, even though no such flaws have been identified by Ultrasonic Test. Consequently, Results show that the predicted flaw sizes will determine acceptability for continued service and maintenance.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.7
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.226-233
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• 1995

The wall thickness of a pressure tube is increased in order to reduce the probability of failure in a pressure tube of CANDU type reactor. It is presented here that the variation of wall thickness changes stress, hydrogen concentration and delayed hydride cracking in Zr-2.5Nb pressure tube. When the wall thickness is increased from 4.2 mm to 5.2 mm, the stress exerted on the tube and the deuterium taken up during operation are reduced by 19％. Further, the calculated allowable depth of the surface flaw over which delayed hydride cracking(DHC) is susceptible increases by 50％. DHC initiation is controlled by the stress and by the hydrogen concentration in the pressure tube. The results are therefore very significant in such a respect that increased wall thickness may reduce DHC initiation. Ac the wall thickness increases the hydrostatic tension will increase. Its impact on the acceleration of the crack growth rate of DHC deserves further studies.