Effects of Fracture Toughness Curves of ASME Section XI: Appendix G on a Reactor Pressure Vessel Under Pressure–Temperature Limit Operation

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.

Effect of non‐uniform reactor cooling on fracture and constraint of a reactor pressure vessel

“…The reactor pressure vessel (RPV) in a nuclear power plant is not replaceable, and thus, its integrity determines the lifetime of the nuclear power plant. During the operation of the plant, the integrity of RPV should be assured, and the brittle failure of the RPV should be excluded . A critical loading for a RPV is pressurized thermal shock (PTS), ie, rapid cooling of sections of the hot and still pressurized RPV by injection of cold emergency coolant, which is the result of the loss‐of‐coolant accidents (LOCAs) .…”

“…Because of the significance and nuclear safety, a lot of work has been done concerning the safety assessment of RPVs under accident scenarios . Both deterministic and probabilistic integrity analyses of an RPV subjected to PTS transient considering uniform RPV cooling have been performed by linear elastic (LEFM) and elastic‐plastic fracture mechanics (EPFM) . A review of the procedures, methods, and computer codes for the RPV integrity assessment is given in Qian and Niffenegger .…”

Effects of embedded cracks and residual stresses on the integrity of a reactor pressure vessel

“…During the last two decades, substantial progress has been achieved in the safety assessment of RPVs under PTS loadings [1][2][3][4][5][6][7][8][9][10][11][12][13]. Integrity analysis of a model RPV subjected to a small break loss-of-coolant accident (SBLOCA) and a medium break loss-of-coolant accident (MBLOCA) is performed by assuming crack depths of two times the nondestructive examination limit [1][2].…”

“…Elastic-plastic analysis and cleavage fracture analysis have been performed in RPV materials considering different loading transients [3][4][5][6][7]. Chou et al [8] performed a probabilistic analysis for a RPV considering the uncertainty of crack distributions and fracture toughness according to the US Nuclear Regulatory Commission [9]. Testing of fracture toughness in the ductile-to-brittle transition temperature is documented in [10].…”

Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

“…WPS is a phenomenon of improvement of apparent fracture toughness of ferritic steels in lower shelf region by preloading them at upper shelf region [11]. Therefore, under some cooling type transients most crack initiation may be excluded when the WPS is considered [12,13], even if is larger than IC . If a flaw is in a state of WPS, it is not eligible for initiation (or reinitiation after the flaw has ever been arrested) until the following conditions are met [9]: (1) of the flaw tip is greater than After a crack initiation, FAVOR assumes that the flaw becomes surface breaking with infinite aspect ratio and then enters the Initiation-Growth-Arrest (IGA) submodel to evaluate the crack propagation and ductile-tearing behavior.…”