PWR pressure vessel integrity during overcooling accidents
Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5869170
- Report Number(s):
- CONF-811042-10; ON: TI85004515
- Resource Relation:
- Conference: 9. water reactor safety research information meeting, Washington, DC, USA, 26 Oct 1981
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FRACTURE MECHANICS
COMPUTER CODES
PRESSURE VESSELS
THERMAL SHOCK
PWR TYPE REACTORS
REACTOR SAFETY
CRACK PROPAGATION
CRACKS
ECCS
LOSS OF COOLANT
PRIMARY COOLANT CIRCUITS
REACTOR ACCIDENTS
RELIABILITY
STRESS ANALYSIS
THERMAL STRESSES
ACCIDENTS
CONTAINERS
COOLING SYSTEMS
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
MECHANICS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
STRESSES
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled