Stress Intensity Factor Solutions for Continuous Surface Flaws in Reactor Pressure Vessels

2009 ◽  
pp. 385-385-18 ◽  
Author(s):  
CB Buchalet ◽  
WH Bamford
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Pressure Vessels ◽  
Surface Flaws ◽  
Continuous Surface ◽  
Reactor Pressure

Stress Intensity Factor Influence Coefficients for External Surface Flaws in Boiling Water Reactor Pressure Vessels

2009 ◽  
Author(s):  
Shengjun Yin ◽  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
External Surface ◽  
Pressure Vessels ◽  
Water Reactor ◽  
Influence Coefficients ◽  
Surface Flaws ◽  
Reactor Pressure

Over the service life of a nuclear power plant, the Boiling Water Reactor (BWR) may undergo many cool-down and heat-up thermal-hydraulic transients associated with, for example, scheduled refueling outages and other normal operational transients, or even possible overcooling transients. These thermal-hydraulic events can act on postulated surface flaws in BWRs and therefore impose potential risk on the structure integrity of Reactor Pressure Vessels (RPVs). Internal surface flaws are of interest for the BWRs under overcooling accidental scenarios, while external surface flaws are more vulnerable when the BWRs are subjected to heat-up transients. Stress Intensity Factor Influence Coefficient (SIFIC) databases for application to linear elastic fracture mechanics analyses of BWR pressure vessels which typically have an internal radius to wall thickness ratio (Ri/t) between 15 and 20 were developed for external surface breaking flaws. This paper presents three types of surface flaws necessary in fracture analyses of BWRs: (1) finite-length external surface flaws with aspect ratio of 2, 6, and 10. (2) infinite-length axial external surface flaws; and (3) 360° circumferential external surface flaws. These influence coefficients have been implemented and validated in the FAVOR fracture mechanics code developed at Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission (NRC). Although these SIFIC databases were developed in application to RPVs subjected to thermal-hydraulic transients, they could also be applied to RPVs under other general loading conditions.

Effect of Cladding Residual Stress Modeling Technique on Shallow Flaw Stress Intensity Factor in a Reactor Pressure Vessel

2015 ◽  
Author(s):  
Joshua Kusnick ◽  
Mark Kirk ◽  
B. Richard Bass ◽  
Paul Williams ◽  
Terry Dickson
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Residual Stresses ◽  
Room Temperature ◽  
Pressure Vessels ◽  
Temperature Model ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

Prior probabilistic fracture mechanics (PFM) analysis of reactor pressure vessels (RPVs) subjected to normal cool-down transients has shown that shallow, internal surface-breaking flaws dominate the RPV failure probability. This outcome is caused by the additional crack driving force generated near the clad interface due to the mismatch in coefficient of thermal expansion (CTE) between the cladding and base material, which elevates the thermally induced stresses. The CTE contribution decreases rapidly away from the cladding, making this effect negligible for deeper flaws. The probabilistic fracture mechanics code FAVOR (Fracture Analysis of Vessels, Oak Ridge) uses a stress-free temperature model to account for residual stresses in the RPV wall due to the cladding application process. This paper uses finite element analysis to compare the stresses and stress intensity factor during a cool-down transient for two cases: (1) the existing stress-free temperature model adopted for use in FAVOR, and (2) directly applied RPV residual stresses obtained from empirical measurements made at room temperature. It was found that for a linear elastic fracture mechanics analysis, the application of measured room temperature stresses resulted in a 10% decrease in the peak stress intensity factor during a cool-down transient as compared to the stress-free temperature model.

Stress Intensity Factor Solutions for Internal Longitudinal Semi-Circular Surface Flaws in a Cylinder Under Arbitrary Loading

1983 ◽  
Vol 105 (4) ◽  
pp. 309-315 ◽  
Author(s):  
Y. S. Lee ◽  
M. Raymund
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Degrees Of Freedom ◽  
Pressure Vessels ◽  
Loading Conditions ◽  
Arbitrary Loading ◽  
Surface Flaws ◽  
Circular Surface ◽  
Inside Radius

The behavior of semi-circular surface flaws in cylinders is of interest in the technology of pressure vessels. The object of this study is to determine the stress intensity factor distribution around the crack front under arbitrary loading conditions for a longitudinal semi-circular flaw with Ri/t = 10; where Ri is the inside radius of the cylinder, and t is the cylinder thickness. Six crack depths are studied under various loading conditions: a/t = 0.10, 0.25, 0.50, 0.65, 0.80, and 0.90, where a is the circular flaw radius. In general, the finite element method is used to determine the displacement solution. Parks’ stiffness derivative method is used to find the stress intensity factor distribution around the semi-circle. The immediate crack tip geometry is modeled by use of a “macro-element” containing over 1600 degrees of freedom. Four separate loadings are considered: 1) constant, 2) linear, 3) quadratic, and 4) cubic crack surface pressure. From these loadings, nondimensional magnification factors are derived to represent the resulting stress intensity factors. Comparisons are made with other investigators and results agree within 5 percent of published results.

scholarly journals Surrogate Model Development and Validation for Reliability Analysis of Reactor Pressure Vessels

2016 ◽  
Author(s):  
William M. Hoffman ◽  
Matthew E. Riley ◽  
Benjamin W. Spencer
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Time History ◽  
Pressure Vessels ◽  
Training Data ◽  
Influence Coefficient ◽  
Parameter Uncertainties ◽  
Benchmark Solution ◽  
Reactor Pressure

In nuclear light water reactors, the reactor core is contained within a thick walled steel reactor pressure vessel (RPV). Over time, material embrittlement caused by exposure to neutron flux makes the RPV increasingly susceptible to fracture under transient conditions. Because of parameter uncertainties, probabilistic methods are widely used in assessing RPV integrity. For efficient probabilistic analysis, techniques to rapidly evaluate the stress intensity factor for given flaw geometry and stress conditions are essential. The stress intensity factor influence coefficient (SIFIC) technique is widely used for this purpose, but is limited to axis-aligned flaw geometries. To consider a wider range of flaw geometries, surrogate models to compute stress intensity factors are explored. Four surrogate modeling techniques are applied here to compute SIFICs from a set of training data, including two different response surface polynomials, a model utilizing ordinary kriging and another using interpolation. Errors in the SIFICs are assessed for all of these techniques. These techniques are benchmarked against a benchmark solution by computing the time history of the stress intensity factor for an axis-aligned, semi-elliptical surface breaking flaw in an RPV subjected to a transient loading history. All of these techniques compare well with the benchmark solution.

Plasticity Correction on Stress Intensity Factor Evaluation for Underclad Cracks in Reactor Pressure Vessels

2020 ◽  
Vol 142 (5) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Base Metal ◽  
Structural Integrity ◽  
Correction Method ◽  
Pressure Vessels ◽  
Integrity Assessment ◽  
Plasticity Effect ◽  
Reactor Pressure

Abstract Structural integrity assessment of reactor pressure vessels (RPVs) is essential for the safe operation of nuclear power plants. For RPVs in pressurized water reactors (PWRs), the assessment should be performed by considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. To assess the structural integrity of an RPV, a traditional method is usually employed by comparing fracture toughness of the RPV material with the stress intensity factor (KI) of a crack postulated near the RPV inner surface. When an underclad crack (i.e., a crack beneath the cladding of an RPV) is postulated, KI of this crack can be increased owing to the plasticity effect of cladding. This is because the yield stress of cladding is lower than that of base metal and the cladding may yield earlier than base metal. In this paper, detailed three-dimensional (3D) finite element analyses (FEAs) were performed in consideration of the plasticity effect of cladding for underclad cracks postulated in Japanese RPVs. Based on the 3D FEA results, a plasticity correction method was proposed on KI calculations of underclad cracks. In addition, the effects of RPV geometries and loading conditions were investigated using the proposed plasticity correction method. Moreover, the applicability of the proposed method to the case which considers the hardening effect of materials after neutron irradiation was also investigated. All of these results indicate that the proposed plasticity correction method can be used for KI calculations of underclad cracks and is applicable to structural integrity assessment of Japanese RPVs containing underclad cracks.

Structural Integrity of Penetration Nozzle of Reactor Pressure Vessel Head of PWR Plant

2008 ◽  
Author(s):  
Kiminobu Hojo ◽  
Naoki Ogawa ◽  
Yoichi Iwamoto ◽  
Kazutoshi Ohoto ◽  
Seiji Asada ◽  
...  
Keyword(s):  
Residual Stress ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Stress Field ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Complex Structure ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

Sign in / Sign up

Export Citation Format

Share Document