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

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Theodore, A. Lang
Copyright Year	2003
Abstract	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.
Sponsorship	Pressure Vessels and Piping Division
Starting Page	169
Ending Page	176
Page Count	8
File Format	PDF
ISBN	0791841596
DOI	10.1115/PVP2003-2167
Volume Number	Aging Management and Component Analysis
Conference Proceedings	ASME 2003 Pressure Vessels and Piping Conference
Language	English
Publisher Date	2003-07-20
Publisher Place	Cleveland, Ohio, USA
Access Restriction	Subscribed
Subject Keyword	Water Cycles Risk Stress corrosion cracking Alloy steel Cavities Welded joints Pressure vessels Nozzles Coolants Damage Cladding systems (building) Leakage Cracking (materials) Construction Reactor vessels Cooling Alloys Inspection Fracture (materials) Fracture (process) Corrosion Stainless steel Nuclear reactors In-service inspection Asme standards Accidents
Content Type	Text
Resource Type	Article

Sl.	Authority	Responsibilities	Communication Details
1	Ministry of Education (GoI), Department of Higher Education	Sanctioning Authority	https://www.education.gov.in/ict-initiatives
2	Indian Institute of Technology Kharagpur	Host Institute of the Project: The host institute of the project is responsible for providing infrastructure support and hosting the project	https://www.iitkgp.ac.in
3	National Digital Library of India Office, Indian Institute of Technology Kharagpur	The administrative and infrastructural headquarters of the project	Dr. B. Sutradhar bsutra@ndl.gov.in
4	Project PI / Joint PI	Principal Investigator and Joint Principal Investigators of the project	Dr. B. Sutradhar bsutra@ndl.gov.in Prof. Saswat Chakrabarti will be added soon
5	Website/Portal (Helpdesk)	Queries regarding NDLI and its services	support@ndl.gov.in
6	Contents and Copyright Issues	Queries related to content curation and copyright issues	content@ndl.gov.in
7	National Digital Library of India Club (NDLI Club)	Queries related to NDLI Club formation, support, user awareness program, seminar/symposium, collaboration, social media, promotion, and outreach	clubsupport@ndl.gov.in
8	Digital Preservation Centre (DPC)	Assistance with digitizing and archiving copyright-free printed books	dpc@ndl.gov.in
9	IDR Setup or Support	Queries related to establishment and support of Institutional Digital Repository (IDR) and IDR workshops	idr@ndl.gov.in

Weld Residual Stresses and Primary Water Stress Corrosion Cracking in Bimetal Nuclear Pipe Welds

U.S. Industry Activities on Inspection of CRDM Penetrations

Flaw Size Acceptance Limits for a Stainless Steel Pressure Vessel

Regulatory Activities Related to Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles

A Study of the Key Parameters Affecting the Design of Optimized Weld Overlays

Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment

ASME Subcommittee XI Initiatives for the Detection and Correction of Boric Acid Corrosion of Pressure Retaining Low Alloy Steel Components in Pressurized Water Reactor Plants

Welding Residual Stress and Flaw Evaluation for Dissimilar Metal Welds With Alloy 52 Inlays

PWSCC Crack Growth Modeling Approaches

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

Similar Documents

Weld Residual Stresses and Primary Water Stress Corrosion Cracking in Bimetal Nuclear Pipe Welds

U.S. Industry Activities on Inspection of CRDM Penetrations

Flaw Size Acceptance Limits for a Stainless Steel Pressure Vessel

Regulatory Activities Related to Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles

A Study of the Key Parameters Affecting the Design of Optimized Weld Overlays

Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment

ASME Subcommittee XI Initiatives for the Detection and Correction of Boric Acid Corrosion of Pressure Retaining Low Alloy Steel Components in Pressurized Water Reactor Plants

Welding Residual Stress and Flaw Evaluation for Dissimilar Metal Welds With Alloy 52 Inlays

PWSCC Crack Growth Modeling Approaches

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