Integrity of PWR pressure vessels during overcooling accidents

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

OSTI ID:5091593

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. For the purpose of evaluating this problem 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 today if subjected to a Rancho Seco (1978) or TMI-2 (1979) 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: 5091593
• Report Number(s): CONF-820802-12; ON: DE82020812
• Country of Publication: United States
• Language: English

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36 MATERIALS SCIENCE
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ALLOYS
CONTAINERS
CRACKS
FAILURES
FRACTURE PROPERTIES
IRON ALLOYS
IRON BASE ALLOYS
MECHANICAL PROPERTIES
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
PWR TYPE REACTORS
RADIATION EFFECTS
REACTORS
STEELS
STRESS ANALYSIS
THERMAL SHOCK
WATER COOLED REACTORS
WATER MODERATED REACTORS