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.

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.

On Corrosion Resistance Of Austenitic Stainless Steel Clad Layer on a Low Alloy Steel

2018 ◽  
Vol 32 (3) ◽  
pp. 20
Author(s):  
Manas Kumar Saha ◽  
Ritesh Hazra ◽  
Ajit Mondal ◽  
Santanu Das
Keyword(s):  
Stainless Steel ◽  
Corrosion Resistance ◽  
Austenitic Stainless Steel ◽  
Alloy Steel ◽  
Low Alloy Steel ◽  
Clad Layer

Characterization of Clad Joints of High Strength Low Alloy Steel with Stainless Steel and Titanium

2018 ◽  
Vol 51 (4) ◽  
pp. 46
Author(s):  
N. Venkateswara Rao ◽  
G. Madhusudhan Reddy ◽  
S. Nagarjuna
Keyword(s):  
Stainless Steel ◽  
Alloy Steel ◽  
High Strength ◽  
Low Alloy Steel

Corrosion Risk and Repassivation of Duplex Stainless Steel UNS S82551 in Treated Seawater Injection Service

CORROSION ◽  
10.5006/3697 ◽  
2021 ◽  
Author(s):  
Nicolas Larche ◽  
Perry Nice ◽  
Hisashi Amaya ◽  
Lucrezia Scoppio ◽  
Charles Leballeur ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Dissolved Oxygen ◽  
Duplex Stainless Steel ◽  
Oxygen Content ◽  
Alloy Steel ◽  
Stainless Steels ◽  
Crevice Corrosion ◽  
Low Alloy Steel ◽  
Corrosion Risk ◽  
Service Conditions

In seawater injection wells, the available well tubing materials are generally Low alloy steel, Glass Reinforced Epoxy lined low alloy steel or Corrosion Resistant Alloy’s (CRA) such as super duplex stainless steel. However, in treated seawater the corrosion risk can be controlled and lower grade alloys (low alloy steel) can be considered. But for long well lifetime designs (20 years plus), then low alloy steel tubing can be challenged. In this respect recent efforts have focused attention on better dissolved oxygen control which permits the investigation and on the possible use of more cost-effective materials such as the duplex stainless steels UNS S82551, and UNS S82541 (the latter is a higher strength version, but same PRENw). Full scale testing of tubes joined together with a proprietary premium threaded connection (PCPC couplings) was performed in controlled seawater loops simulating service conditions at 30°C. The flow rate and dissolved oxygen were controlled at 5 m/s and <20ppb, respectively. Weekly dissolved oxygen excursions corresponding to 24h at 100ppb followed by 1 hour at 300ppb were performed during the 5 months exposure. Corrosion results of UNS S82551/S82541 tubing were compared to UNS S31803 and UNS S39274. In parallel, laboratory exposures of creviced coupons for parametric study were performed in dissolved oxygen-controlled cells, allowing the measurement of electrochemical potentials as function of dissolved oxygen content and the related corrosion resistance. The results showed that dissolved oxygen content should be properly controlled below critical values to avoid crevice corrosion of the lesser alloyed duplex stainless steels. The ability of UNS S82541 to recover or re-passivate after prolonged exposures to high dissolved oxygen concentrations (DOC) was also determined with both the use of full sized pipe-coupling premium connection (PCPC) test cells, and electrochemical testing involving a Remote Crevice Assembly (RCA). The re-passivation potential was investigated after different active crevice corrosion durations. The results of the study allowed to precisely define the limits of use of UNS S82541 in treated seawater, i. e. the critical DOC conditions for corrosion initiation and for re-passivation of UNS S82541. For all tested conditions, the UNS S82551/S82541 showed a rather good ability to re-passivation when normal service conditions (i. e. low dissolved oxygen) are recovered.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

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