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Research Papers: Design and Analysis

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

[+] Author and Article Information
Kuan-Rong Huang

National Chung-Shan Institute of Science
and Technology,
No. 481, 6th Neighborhood, Sec. Jia'an,
Zhongzheng Rd., Longtan Dist.,
Taoyuan City 32546, Taiwan
e-mail: d94543005@ntu.edu.tw

Chin-Cheng Huang

Institute of Nuclear Energy Research,
No. 1000, Wenhua Rd.,
Jiaan Village, Longtan Township,
Taoyuan County 32546, Taiwan
e-mail: cchuang@iner.gov.tw

Hsiung-Wei Chou

Institute of Nuclear Energy Research,
No. 1000, Wenhua Rd.,
Jiaan Village, Longtan Township,
Taoyuan County 32546, Taiwan
e-mail: hwchou@iner.gov.tw

1Corresponding author.

Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received May 4, 2015; final manuscript received September 28, 2015; published online February 8, 2016. Assoc. Editor: Kunio Hasegawa.

J. Pressure Vessel Technol 138(3), 031202 (Feb 08, 2016) (9 pages) Paper No: PVT-15-1089; doi: 10.1115/1.4032110 History: Received May 04, 2015; Revised September 28, 2015

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

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References

Figures

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Fig. 1

Illustration comparison of aKIc and ASME KIc

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Fig. 2

The configuration of RPV beltline region

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Fig. 3

Time histories of temperature and pressure for the LOCA event

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Fig. 4

Time histories of temperature and pressure for the MSLB event

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Fig. 5

Time histories of temperature and pressure for loss-of-power event 1

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Fig. 6

Time histories of temperature and pressure for loss-of-power event 2

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Fig. 7

Time histories of temperature and pressure for the PTS event

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Fig. 8

Time histories of temperature and pressure for the SGTR event

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Fig. 10

Time histories of the applied KI and relative temperature, T − RTNDT, at the crack tip in the case of the RPV at 54 EFPY subjected to the loss-of-power event

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Fig. 9

Time histories of the applied KI and relative temperature, T − RTNDT, at the crack tip in the case of the RPV at 54 EFPY subjected to the LOCA event

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Fig. 11

Time histories of the applied KI and relative temperature, T − RTNDT, at the crack tip in the case of the RPV at 54 EFPY subjected to the PTS event

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Fig. 12

Time histories of the applied KI and relative temperature, T − RTNDT, at the crack tip in the case of the RPV at 54 EFPY subjected to the SGTR event

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Fig. 13

Illustration of the applied KI time history and the Weibull KIc statistical model considering different RTNDTs in the case of the RPV at 54 EFPY subjected to the PTS event

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Fig. 14

Illustration of the applied KI time history and the Weibull KIc statistical model considering different RTNDTs in the case of the RPV at 54 EFPY subjected to the PTS even

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