The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

The objective of this study is to investigate the feasibility of simulating the mechanical properties of irradiatied austenitic stainless steels by cold-working. In this study, the tensile properties, cyclic hardening behaviors and fracture toughness of cold-worked TP316L stainless steel were compared with those of austenitic stainless steels irradiated by neutrons. It showed that cold-working can properly simulate the increase in strength and the decrease in ductility and fracture resistance of austenitic stainless steels by neutron irradiation, even though it could not perfectly simulate the microstructures of irradiated austenitic stainless steels. Also, cold-working can appropriately simulate the hardening behaviors of neutron irradiated austenitic stainless steels under monotonic and cyclic loading conditions.

In this study, to evaluate conservatism of various fatigue life evaluation procedures, fatigue tests were conducted using compact tension (CT) specimens with a round notch, made of A516 Gr.70 carbon steel and A240 TP304 stainless steel, under load-controlled cyclic condition. Experimental fatigue failure cycles were measured and compared with predicted fatigue lives using two different life evaluation methods; (1) Design-By-Analysis (DBA) procedure given in ASME B&PV Code, Sec. III, Div. 1, Subsec. NB-3200 and (2) structural stress-based approach provided in ASME B&PV Code, Sec. VIII, Div. 2, Part 5. To predict fatigue failure cycles, three-dimensional elastic finite element analysis was conducted. Fatigue lives were predicted by both design fatigue curve given in ASME B&PV Code, Sec. III, Div. 1, Appendices and best-fit fatigue curve suggested in NUREG/CR-6815 for the DBA procedure. Finally, fatigue lives evaluated by various methods were compared with test results, and then conservatism between each evaluation procedure was discussed.

Cracked component analysis is needed for structural integrity analysis under seismic loading. Under large amplitude cyclic loading conditions, the change in material properties can be complex, depending on the magnitude of plastic strain. Therefore the cracked component analysis under cyclic loading should consider appropriate cyclic hardening model. This study introduces a procedure for determining an appropriate cyclic hardening model for cracked component analysis. The test material was nuclear-grade TP316 stainless steel. The material cyclic hardening was simulated using the Chaboche combined hardening model. The kinematic hardening model was determined from standard tensile test to cover the high and wide strain range. The isotropic hardening model was determined by simulating C(T) test under cyclic loading using ABAQUS debonding analysis. The suitability of the material hardening model was verified by comparing load-displacement curves of cyclic C(T) tests under different load ratios.

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

This paper investigates deformation characteristics of gas pipeline using the in-plane bending experiment and finite element analysis of a pipe bend. The effect of the bending angle and internal pressure on the deformation characteristics is analyzed. The pipe bend used in this study is API 5L X65 (out diameter: 20 inch) material with the thickness of 11.9 mm. The maximum load, displacement at maximum load, angle and local strain of 90° pipe bend are obtained from the in-plane bending experiment. Comparison between FE results and experimental data shows overall good agreements. In addition, the deformation characteristics of 22.5° and 45° pipe bend are calculated using the finite element analysis. As a result, the effect of the bend angle on the deformation characteristics is discussed.

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

In this study, limit load equations of thick-walled pipes with inner and outer circumferential surface cracks were derived based on force and moment equilibrium conditions. Since the limit load equations based on the mean radius at uncracked ligament, previously proposed by Kanninen et al., are based on the premise that the pipe wall thickness is relatively thin, the existing limit load solutions are only applicable to thin-walled pipes. In order to analyze the effect of the pipe thickness and surface crack depth on the limit load results, the predictions using the present limit load equations are compared with those using the existing solutions for thin-walled pipes. Being derived considering the thickness effect, the limit load solutions from this paper are believed to be more accurate for thick-walled pipes than the limit load equations presented for thin-walled pipes, and thus to be valuable equations for integrity assessment of thick-walled pipes.