Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

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.

Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

“…Details on these PTS events including time histories of coolant temperature and inner pressure, and occurrence frequencies are provided in the NUREG report (EricksonKirk et al, 2006). Table 2 Crack Described in Table 4 PTS event 13 events from the NUREG report (EricksonKirk et al, 2006) ·SBLOCA #3, #114 ·MBLOCA #7 ·LBLOCA #56 ·MSLB #103 ·SOV #60, #71, #94, #97, #123, #126, #129, and #130 Neutron fluence ・Mean value of maximum neutron fluence: 7 ×10 19 n/cm2 (E > 1 MeV) ・SD: 13.1% of the mean value ・Spatial variation in the core region: based on fluence map in a previous paper (Chou and Huang, 2014) Operating temperature Same as that in Table 2 Initial RTNDT ・Mean value: 5ºC for base metal and 50 ºC for weld ・SD:…”

Study on Thermal Shock Failure Characteristics of RPV During IVR System Start-Up for Typical Advanced PWR

“…The stress intensity factors for the underclad and through clad defects were studied in a reactor pressure vessel subjected to a pressurized thermal shock (Marie et al, 2005). There are also many key impact studies on the PTS, including elastic and elastoplastic fracture analyses (Sun et al, 2017), constraint effect analysis (Kim et al, 2003), and structural reliability evaluation (Chou and Huang, 2014). Therefore, it is necessary to carry out thermal shock research for the HPR1000 nuclear power plant in order to study the safety assessment of the PTS of the RPV.…”

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

“…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). [1][2][3][4][5]13,[16][17][18][19][20][21][22] A review of the procedures, methods, and computer codes for the RPV integrity assessment is given in Qian and Niffenegger. 18 Probabilistic assessment is performed for an RPV subjected to small break LOCA (SBLOCA) and medium break LOCA (MBLOCA) transients.…”

“…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 .…”