Integrity of PWR pressure vessels during overcooling accidents

Conference · Fri Jan 01 04:00:00 EST 1982

OSTI ID:6627313

Cheverton, R D; Iskander, S K; Whitman, G D

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 6627313
• Report Number(s): CONF-821037-27; ON: DE83001936
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
CONTAINERS
FAILURES
FRACTURE MECHANICS
MECHANICS
PRESSURE VESSELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
STRESSES
THERMAL SHOCK
THERMAL STRESSES
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS