Steam generator tube failures

Technical Report · Mon Apr 01 04:00:00 EST 1996

DOI:https://doi.org/10.2172/236258· OSTI ID:236258

MacDonald, P E; Shah, V N; Ward, L W; Ellison, P G

A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Programs; Idaho National Engineering Lab., Idaho Falls, ID (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 236258
• Report Number(s): NUREG/CR--6365; INEL--95/0383; ON: TI96010605
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CALHOUN-1 REACTOR
CANDU TYPE REACTORS
DOEL-2 REACTOR
FAILURES
GINNA-1 REACTOR
IN-SERVICE INSPECTION
MC GUIRE-1 REACTOR
MIHAMA-1 REACTOR
NONDESTRUCTIVE TESTING
NORTH ANNA-1 REACTOR
PALO VERDE-2 REACTOR
POINT BEACH-2 REACTOR
PRAIRIE ISLAND-1 REACTOR
PWR TYPE REACTORS
RISK ASSESSMENT
SERVICE LIFE
STEAM GENERATORS
SURRY-2 REACTOR
TUBES
WWER TYPE REACTORS