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Structural integrity analysis for reactor pressure vessel (RPV) nozzles has been a vital issue due to the irradiation embrittlement and especially the discontinuity of geometry in RPV nozzles, which may cause a higher stress concentration and influence the structural integrity of RPV nozzles. For the linear elastic fracture mechanics (LEFM) analysis of the RPV nozzle, a postulated crack with a specific crack depth is presumed to evaluate the stress intensity factors (SIFs) whose crack depth does not grow during LEFM analysis. Thus, this paper would discuss the crack propagation for the postulated crack with a specific crack depth under the different reference temperature of nil-ductility transition (R T N D T ), and the SIFs on the specific crack depth are also calculated. In recent research, the extended finite element method (XFEM) was developed, which is based on the enrichment solution space to simplify the computational fracture mechanics with the crack propagation of various discontinuities, and can be used to evaluate the crack propagation of RPV nozzle with a specific crack depth under R T N D T of −23.89, −10.89, 0.89, 2.69, and 2.89 °C. To investigate the crack propagation and structural integrity of RPV nozzles, the finite element analysis of Taiwan domestic boiling water reactor (BWR) nozzles containing the postulated circular corner crack with depths of one-fourth and one-tenth thickness along the 45° path from the tip of the nozzle corner is engaged. The present result shows that the crack was growing when the R T N D T was 0.89 °C and. In addition, the XFEM calculation of the SIFs for the specific crack depth is similar to the LEFM results. • The RPV nozzles with postulated cracks were established to investigate crack propagation through XFEM. • The differences in RT NDT were used to evaluate the fracture toughness and discuss the crack propagation for RPV nozzles. • The stress intensity factors (SIFs) obtained by XFEM, LEFM, and the ASME BPVC were compared. • The XFEM analysis would overestimate SIFs more than the ASME BPVC and the LEFM. [ABSTRACT FROM AUTHOR]

Abstract: The paper describes the computation of the maximum horizontal stress (σ H) magnitude in the vicinity of the Chelungpu Fault, Taiwan, host of the slip zone during the Chi-Chi earthquake (M w 7.6; 1999). The scientific hole B intercepts the Chelungpu Fault at 1136 m. At the depths of logged breakouts (940–1310 m), the vertical stress (σ v) as estimated from density logs increases linearly with depth from 22 to 31 MPa. A series of leak-off tests yielded two reliable shut-in pressures, 23.7 MPa at 1085 m and 29.8 MPa at 1279 m, which are lower than the estimated σ v, albeit by only 2.1 and 0.6 MPa, respectively. In our analysis the shut-in pressures were considered to represent estimates of the least horizontal principal stresses (σ h) at the respective depths, and consequently the test-induced fractures were assumed to have been vertical. Principal stress directions had been previously determined by others (105°–155° for the maximum horizontal stress, σ H, except in the immediate vicinity of the Chelungpu Fault). The contribution of this paper is the estimation of the σ H magnitude by considering that the state of stress at the points of intersection between breakout and borehole wall is in a state of limit equilibrium with the true triaxial strength criterion. The resulting σ H in the range of logged breakouts increases with depth from 55 MPa at 940 m to 59 MPa at 1310 m. Thus, the estimated state of stress prevailing across the Chelungpu Fault is compatible with strike-slip, but marginally also with thrust faulting. However, the likelihood that the shut-in pressures actually represent σ v magnitudes, and that the leak-off test-induced fractures were sub-horizontal, cannot be ignored. In that case the state of stress would clearly favor thrust faulting. [Copyright &y& Elsevier]

• The ASME formula used for calculating stress intensity factor for nozzles were verified by linear elastic fracture mechanics finite element analysis. • Five Taiwan domestic RPV nozzles were considered for finite element stress and fracture mechanics analysis. • ASME formula provides conservative results for circular-shape nozzle corner crack but underestimates for elliptical-shaped crack. • The adjusted magnification factors were proposed for the elliptical-shaped nozzle corner crack. Due to the structural discontinuity of geometry having higher stress concentration, the nozzles may dominate the structural integrity of reactor pressure vessel (RPV) system, even the material radiation embrittlement is insignificant. The nonmandatory Appendix G in Section XI of the ASME Boiler and Pressure Vessel code provides a fracture mechanics-based methodology using magnification factors to calculate the stress intensity factor of reference crack tips as long as the stress distribution through the nozzle section is available. This paper establishes various three-dimensional finite element models of Taiwan domestic RPV nozzles to conduct a series of stress and linear elastic fracture mechanics analyses to estimate the stress intensity factors. The circular and elliptical reference corner cracks with depths of one-fourth and one-tenth thickness of the vessel wall along the 45° path are considered. Two evaluation results using the ASME method and the finite element linear elastic fracture mechanics analysis are compared. It is found that the ASME method produces conservative values of stress intensity factor at crack tips of circular corner cracks but underestimates the elliptical corner cracks. Finally, an adjusted magnification factor formula is proposed to obtain more accurate stress intensity factors for elliptical-shaped nozzle corner cracks. [ABSTRACT FROM AUTHOR]

• PFM method is used to evaluate the fracture risk of decommissioning BWR vessels subjected to seismic loads. • The manner using equivalent internal pressure to simulate the driving forces caused by seismic loads is proposed. • Circumferential shell welds dominate the fracture risk for a shutdown RPV. • Present result demonstrates the feasibility of ISI relief on shell welds of RPV during decommissioning transition period. During decommissioning transition period, the hazards from internal pressurization and thermal transients to a reactor vessel can be eliminated, but the effects of seismic loads on vessel integrity may be of concern. In this paper, we employed the Oak Ridge National Laboratory's (ORNL) FAVOR code to perform probabilistic integrity assessment for the Taiwan domestic boiling water reactor (BWR) vessels of a decommissioning plant considering seismic loads. A manner to transform seismic loads as the equivalent internal pressure for FAVOR analysis was proposed. Two probabilistic fracture mechanics (PFM) models, the conservative model that U.S. NRC has used for BWR regulation, and the realistic model reflecting the entire vessel beltline configuration, were analyzed and compared. It is found that the circumferential welds dominate the fracture risk but still can be negligible. Present results could be the reference for weld inspection relief application of the RPV during the decommissioning transition period in Taiwan. [ABSTRACT FROM AUTHOR]

A critical relation for mobilization (failure) of an inclined, uniform soil layer is developed, based on the requirement for balancing the driving stress and the resistance stress acting on the soil layer. Taking into consideration of the uncertainties of parameters of the considered soil layer, the equation of the failure probability (Eq. 18) and the equation of the critical slope at various probabilities of failure (Eq. 22) for the soil layer by applying the first-order second-moment method (FOSM) were developed. The failure probabilities at various mean relative water depths subject to the upper and lower values of the coefficients of variations of parameters and two inclined angles of the soil layers are also estimated by using the Monte Carlo simulation method (MCSM) as well as by using FOSM for comparisons. The results obtained by FOSM are in good agreement with that obtained by MCSM, but the former is more efficient. Given the mean values and the coefficients of variation of related parameters of the soil layer, one can easily estimate the failure probability and the critical failure slope of the soil layer by applying the equations or graphs developed in the present study. [ABSTRACT FROM AUTHOR]

The pressure-temperature limits (P-T limits) of embrittled reactor pressure vessels (RPVs) should be determined for normal heat-up and cool-down operations based on fracture mechanics to ensure the structural integrity. In general, the P-T limits are mainly determined by the fracture toughness of beltline region material which suffers the highest level of neutron embrittlement. However, some other components such as nozzles which suffer insignificant neutron embrittlement may restrict the P-T limits as a result of higher stress concentration from their structural discontinuities. Therefore, beltline region material with the highest reference temperature as well as nozzles that have structural discontinuities need to be considered when developing the P-T limits of RPV. In this paper, the P-T limit calculations for a Taiwan domestic pressurized water reactor (PWR) vessel were performed considering both beltline shell and extended beltline nozzles according to the nonmandatory Appendix G to ASME Code Section XI. The three-dimensional finite element analyses were conducted to obtain the pressure and thermal stress distributions of extended beltline nozzles for P-T limits calculation. It is found that the beltline region still dominates the cool-down P-T limit, but the extended beltline region will partially control the heat-up limit. Further, the reference temperature ranges of extended beltline nozzles that may influence the P-T limits of RPV were also investigated. Present study is helpful for safe operation and also provides a reference for regulation of PWR plants in Taiwan. [ABSTRACT FROM AUTHOR]

Pressurized-thermal-shocks (PTSs) are the most severe accidents that impact the structural integrity of a pressurized water reactor pressure vessel (RPV). In general, the one-dimensional thermal-hydraulic analysis results for PTS transients were often used as the loading conditions for structural integrity evaluation of RPV, but may cause over-conservative results. This paper investigates the effects of the cold plume, which is the non-uniform temperature distribution caused by colder water injected separately into the RPV, on the fracture probability of a three loop pressurized water reactor (PWR) vessel subjected to PTS events. The three-dimensional computational fluid dynamics (CFD) technique associated with probabilistic fracture mechanics (PFM) analysis were employed to comprehensively evaluate the structural integrity of RPV under hypothetical PTS accidents. Firstly the model of a PWR pressure vessel in Taiwan was constructed to simulate the thermal-hydraulic phenomena by CFD methodology. Based on the Best Practice Guidelines (BPGs) and ASME V&V guidance, the estimation of mesh uncertainty has been conducted. Then the detailed thermal-hydraulic boundary conditions corresponding to various RPV shell regions were regarded as the loading conditions for PFM analysis. Present results are compared with those which consider the simplified thermal hydraulic analysis system code's results as boundary conditions. [ABSTRACT FROM AUTHOR]