Erratum: “Factors Affecting the Crack Growth Rates of Reactor Pressure Vessel Steels Under Simulated Boiling Water Reactor Conditions” (Vol. 62, No. 5, p. 403–418)

CORROSION ◽  
2006 ◽  
Vol 62 (6) ◽  
pp. 552-552
Keyword(s):  
Crack Growth ◽  
Growth Rates ◽  
Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Factors Affecting ◽  
Water Reactor ◽  
Pressure Vessel Steels ◽  
Crack Growth Rates ◽  
Reactor Pressure

Radiation Effects in Boiling Water Reactor Pressure Vessel Steels

1980 ◽  
Vol 102 (2) ◽  
pp. 177-186
Author(s):  
J. N. Kass ◽  
A. J. Giannuzzi ◽  
D. A. Hughes
Keyword(s):  
Pressure Vessel ◽  
Radiation Effects ◽  
Statistical Evaluation ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Shelf Energy ◽  
Water Reactor ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Effect of neutron irradiation on notch toughness properties of Boiling Water Reactor pressure vessel steels was determined. Samples from several heats of plate, weld metal, and forgings were irradiated to three different fluence levels and tested. A statistical evaluation of the data was conducted to determine regression analysis mean decreases in upper shelf energy and increases in transition temperature versus fluence.

Cyclic Crack Growth Behavior of Reactor Pressure Vessel Steels in Light Water Reactor Environments

1986 ◽  
Vol 108 (1) ◽  
pp. 26-30 ◽  
Author(s):  
W. A. Van Der Sluys ◽  
R. H. Emanuelson
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Crack Growth ◽  
Growth Rates ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Loading Frequency ◽  
Water Reactor ◽  
Pressure Vessel Steels ◽  
Crack Growth Rates

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533B-1 steels in simulated 550°F Boiling Water Reactor (BWR) and 550°F Pressurized Water Reactor (PWR) environments. Areas investigated over the course of the test program included the effects of loading frequency and R ratio (Kmin/Kmax) on crack growth rate as a function of the stress intensity factor (ΔK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by ΔK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The Higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were higher in a 550°F simulated BWR environment than in a 550°F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency. Further, the loading frequency at which the highest cyclic crack growth rate was observed was found to be a function of the applied ΔK level. In most cases, all cyclic crack growth rates were on or under the ASME Section XI high R water reference flaw growth line and above the Section XI air reference flaw growth line, supporting the position of these lines on the growth rate–ΔK level graph.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

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