No abstract

After the Code Case N-640 was issued in 1999, the fracture toughness curve of reactor pressure vessel materials in ASME Section XI-Appendix G was amended to the KIC curve. In Taiwan, the present pressure-temperature limit curves of normal reactor startup (heat-up) and shut-down (cool-down) for the reactor pressure vessel is still calculated per KIA curve in 1998 or earlier editions. In this paper, the failure risks of a Taiwan domestic reactor pressure vessel under various pressure-temperature limit operations are analyzed. First, the pressure-temperature limit curves of the Taiwan domestic reactor pressure vessel based on KIA and KIC curves, and various levels of embrittlement, are calculated. Then, the ORNL’s probabilistic fracture mechanics code, FAVOR, and the PNNL’s flaw model are utilized to assess the failure probabilities of the reactor pressure vessel under such pressure-temperature limit transients. Further, the deterministic analyses of FAVOR code are also conducted. It is found that under the pressure-temperature limit transients based on KIC curves, the reactor pressure vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the pressure-temperature limits can either increase the operational margin or still maintains the sufficient stability of the analyzed reactor pressure vessel.

No abstract

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

No abstract