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Design and Analysis

Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics (PVP2010-25195)

[+] Author and Article Information
Bo-Yi Chen

e-mail: bychen@iner.gov.tw

Hsien-Chou Lin

Engineer Institute of Nuclear Energy Research,
Taoyuan, Taiwan, R.O.C.

1Corresponding author.

Contributed by the Pressure Vessel and Piping Division of ASME for publication in the Journal of Pressure Vessel Technology. Manuscript received January 3, 2012; final manuscript received June 7, 2012; published online December 17, 2012. Assoc. Editor: Somnath Chattopadhyay.

J. Pressure Vessel Technol 135(1), 011206 (Dec 17, 2012) (4 pages) Paper No: PVT-12-1003; doi: 10.1115/1.4007292 History: Received January 03, 2012; Revised June 07, 2012

The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast neutron fluence exposure. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture analysis of vessels—Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, alloying elements, neutron fluence, and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08 has the probability of failure. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region and provide the aging analysis results for the life extension and the license renewal applications.

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References

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Figures

Grahic Jump Location
Fig. 1

Flowchart and modules of FAVOR: (a) FAVLoad, (b) FAVPFM, and (c) FAVPost

Grahic Jump Location
Fig. 2

The rollout of the domestic reactor pressure vessel

Grahic Jump Location
Fig. 3

The normalized neutron fluence distribution in the beltline region

Grahic Jump Location
Fig. 4

The pressure and temperature of normal and upset transients: (a) Design hydro test, (b) loss of feedwater heaters, and (c) turbine generator trip with isolation valves staying open

Grahic Jump Location
Fig. 5

The pressure and temperature of emergency transients: (a) Loss of feedwater pumps, (b) reactor overpressure, (c) single relief/safety valves blown down, and (d) improper startup

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