Experience on ultrasonic inspection of control rod nozzles of pressurised water reactor pressure vessel

2007 ◽  
Vol 2 (2) ◽  
pp. 129-132
Author(s):  
P. Kauppinen ◽  
H. Jeskanen ◽  
R. Paussu ◽  
B. Elsing
Keyword(s):  
Pressure Vessel ◽  
Ultrasonic Inspection ◽  
Reactor Pressure Vessel ◽  
Control Rod ◽  
Water Reactor ◽  
Reactor Pressure

Significant Corrosion of the Davis-Besse Nuclear Reactor Pressure Vessel Head

2003 ◽  
Author(s):  
Theodore A. Lang
Keyword(s):  
Stainless Steel ◽  
Alloy Steel ◽  
Pressure Vessel ◽  
Human Performance ◽  
Ultrasonic Inspection ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Low Alloy Steel ◽  
Control Rod ◽  
Reactor Pressure

In March of 2002, significant corrosion of the Davis-Besse reactor head was discovered. The Davis-Besse reactor head is of standard construction, composed of low alloy steel and clad with stainless steel. Alloy 600 control rod nozzles penetrate the reactor head, attached with J-groove welds. During an ultrasonic inspection, three of these nozzles were found to have through-wall cracks induced by Primary Water Stress Corrosion Cracking (PWSCC). Undiscovered leakage of borated water over the course of several operating cycles from one of these nozzles led to localized cooling and wastage of the reactor head near the nozzle. This leakage, less than 0.2 gpm (0.8 l/min), was small in comparison to allowable unidentified leakage, but larger than typical PWSCC leakage. The greatest damage to the low alloy steel reactor pressure vessel head was an oblong cavity, approximately 7 × 5 inches (18 × 13 cm), penetrating to the stainless steel cladding. The cracks in this nozzle were axially oriented, which would previously have been considered low risk because they would not have caused control rod ejection. However, the damage led to an increase in risk of a loss of coolant accident, prolonged loss of generation, and replacement of the reactor pressure vessel head. In addition to the industry wide regulatory impact of this event, the Nuclear Regulatory Commission has indicated that there may be a need to revise the inservice inspection requirements in Section XI of the ASME Code. This paper provides a brief synopsis of PWSCC in Control Rod Drive Mechanism nozzles, describes the inspection activities that led to the discovery of both the cracking and the corrosion, and describes the extent and technical cause of the damage. Management and human performance issues that allowed the damage to progress to an advanced state are discussed, since this event would not have been noteworthy if administrative controls and programs had been properly implemented.

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.

Sign in / Sign up

Export Citation Format

Share Document