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TECHNICAL PAPERS

A Finite Element Study on the Integrity Evaluation Method of Subclad Cracks Under Pressurized Thermal Shock Transients

[+] Author and Article Information
Jin-Su Kim, Bon-Geol Koo, Jae-Boong Choi, Young-Jin Kim

School of Mechanical Engineering, Sungkyunkwan University, Kyonggi-do 440-746, Korea

Yun-Won Park

Korea Institute of Nuclear Safety, Daejeon 305-338, Korea

J. Pressure Vessel Technol 125(1), 46-51 (Jan 31, 2003) (6 pages) doi:10.1115/1.1498846 History: Received January 08, 2002; Revised June 05, 2002; Online January 31, 2003
Copyright © 2003 by ASME
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References

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Kwak,  D. O., Choi,  J. B., Kim,  Y. J., Pyo,  C. R., and Park,  Y. W., 1999, “Deterministic Fracture Analysis of Nuclear Reactor Vessel for Pressurized Thermal Shock Accident,” Trans. KSME, 23, No. 8, pp. 1425–1434.
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Keim, E., Hertlein, R., Fricke, S., Schöpper, A., Ternon-Morin, F., and Bezdikian, G., 1999, “Thermal Hydraulics and Fracture Mechanics Analysis of a Small Break Loss of Coolant Accident in the French CP0-Type Reactor Pressure Vessel Integrity Assessment,” Proc., ASME Pressure Vessels and Piping Conference, 388 , pp. 71–77.
ASME Boiler and Pressure Vessel Code, Section XI, 1995, Rules for In-Service Inspection of Nuclear Power Plant Components.
Bass, B. R., 1996, CSNI Project for Fracture Analysis of Large scale International Reference Experiments (FALSIRE II), USNRC Report NUREG/CR-6460.
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Choi,  S. N., Jang,  K. S., Kim,  J. S., Choi,  J. B., and Kim,  Y. J., 2000, “Effect of Cladding on the Stress Intensity Factors in the Reactor Pressure Vessel,” Nucl. Eng. Des., 199, pp. 101–110.
Lee, S. M., Choi, J. B., Kim, Y. J., Park, Y. W., Jhung, M. J., and Pyo, C. R., 2000, “Elastic-Plastic Finite Element Analyses for the Integrity Evaluation of Nuclear Reactor Vessel under Pressurized Thermal Shock Transients,” Proc. ASME Pressure Vessels and Piping Conference, 403 , pp. 79–85.
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ABAQUS Version 5.8, 1999, Hibbitt, Karlsson and Sorensen, Inc.
Korea Institute of Nuclear Safety, 2000, Round Robin Analysis of Pressurized Thermal Shock for Reactor Pressure Vessel, KINS/AR-696.
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Figures

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A schematic illustration of subclad crack
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A schematic illustration of the maximum allowable RTNDT calculation
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Temperature, heat transfer coefficient and pressure distribution for MSLB
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A typical finite element mesh for a/t=1/4
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Resulting stress intensity factors for a/t=1/20
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Comparison of hoop stress distribution between EFEA and EFA after 700 s
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Resulting stress intensity factors for a/t=1/10
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The elastic hoop stress distribution after 700 s
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The elastic hoop stress distribution after 4000 s
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Resulting stress intensity factors for a/t=1/4
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Comparison of maximum allowable RTNDT (maximum criteria)
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Comparison of maximum allowable RTNDT (tangent criteria)

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