scholarly journals EVALUATION OF GALVANIC CORROSION BEHAVIOR OF SA-508 LOW ALLOY STEEL AND TYPE 309L STAINLESS STEEL CLADDING OF REACTOR PRESSURE VESSEL UNDER SIMULATED PRIMARY WATER ENVIRONMENT

2012 ◽  
Vol 44 (7) ◽  
pp. 773-780 ◽  
Author(s):  
Sung-Woo Kim ◽  
Dong-Jin Kim ◽  
Hong-Pyo Kim
Keyword(s):  
Stainless Steel ◽  
Alloy Steel ◽  
Corrosion Behavior ◽  
Pressure Vessel ◽  
Galvanic Corrosion ◽  
Water Environment ◽  
Reactor Pressure Vessel ◽  
Low Alloy Steel ◽  
Primary Water ◽  
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.

Component Reliability Considerations for New Designs and Extended Operation of Boiling Water Reactor (BWRs)

2008 ◽  
Author(s):  
E. Kiss
Keyword(s):  
Stainless Steel ◽  
Alloy Steel ◽  
High Reliability ◽  
Pressure Vessels ◽  
Operational Experience ◽  
Low Alloy Steel ◽  
Irradiation Embrittlement ◽  
Component Reliability ◽  
Extended Operation ◽  
Reactor Pressure

To achieve high reliability for new designs and extended operation of Reactor Pressure Vessels and Internals it is mandatory to apply the technical knowledge gained during operation of the existing Plants to assure that sufficient “Margin” is built into the new design. This paper discusses the importance of four key structural degradation mechanisms that have been shown by operational experience to affect the reliability of the BWR. These are: 1) Stress Corrosion Cracking (IGSCC) of Stainless Steel and Nickel-based Alloys; 2) Irradiation Assisted SCC (IASCC) of Stainless Steel and Nickel-based Alloys; 3) Irradiation Embrittlement of RPV low alloy Steel; 4) Corrosion Assisted Fatigue of Carbon and Low Alloy Steel. While the focus of this paper is the BWR, the mechanisms discussed are equally applicable to the PWR, although the water chemistry effects and mitigations will be different.

Sign in / Sign up

Export Citation Format

Share Document