Comparison of linear-elastic-plastic, elastic-plastic, and fully plastic failure models in the assessment of piping integrity
A double-ended guillotine break in the primary coolant loop of a pressurized water reactor (PWR) is a postulated loss of coolant accient which can result in extreme dynamic loads (i.e., the asymmetric blowdown load) on the reactor pressure vessel (RPV) an vessel intervals. Design and construction of the RPV and support systems to withstand these extreme dynamic loads is very difficult. Similar high loading would also be experienced in a boiling water reactor given a similar accident. Although such a break would be an extremely rare event, its obvious safety and design implications demand that it is carefully evaluated. The work discussed here is part of the Load Combinations Program at Lawrence Livermore National Laboratory to estimate the probability of a double-ended guillotine break in the primary reactor coolant loop of a selected PWR. The program employs a fracture mechanics based fatigue model to propagate cracks from an initial flaw distribution. It was found that while most of the large cracks grew into leaks, a complete (or nearly complete) circumferential crack could lead to a double-ended pipe break with prior leaking and thus, without warning. It is important to assess under what loads such a crack will result in complete pipe severance. The loads considered in this evaluation result from pressure, dead weight and seismic stresses. For the PWR hot leg considered in this investigation the internal pressure contributes the most to the load controlled stresses (i.e., stresses which can cause piping failure) and thus, the problem is treated as axisymmetric with uniform axial loading.
- Research Organization:
- Lawrence Livermore National Lab., CA (USA)
- DOE Contract Number:
- W-7405-ENG-48
- OSTI ID:
- 6353384
- Report Number(s):
- UCRL-84182; CONF-810801-30; ON: TI85016685
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COOLING SYSTEMS
ELASTICITY
ENERGY SYSTEMS
ENERGY TRANSFER
FAILURES
FLUID MECHANICS
FLUID-STRUCTURE INTERACTIONS
GEOMETRY
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATERIALS
MATHEMATICAL MODELS
MATHEMATICS
MECHANICAL PROPERTIES
MECHANICS
PIPES
PRIMARY COOLANT CIRCUITS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR MATERIALS
REACTOR SAFETY
REACTORS
RISK ASSESSMENT
SAFETY
SEISMIC EFFECTS
STRESS ANALYSIS
STRESSES
TENSILE PROPERTIES
ULTIMATE STRENGTH
WATER COOLED REACTORS
WATER MODERATED REACTORS