Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 3. Nonseismic stress analysis. Load Combination Program, Project I final report

Technical Report · Mon Jun 01 04:00:00 EDT 1981

DOI:https://doi.org/10.2172/5320559· OSTI ID:5320559

Chan, A. L.; Lu, S. C.; Rybicki, E. F.; Curtis, D. J.

This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations. 13 refs., 16 figs., 11 tabs.

• Research Organization: Lawrence Livermore National Lab., CA (USA)
• DOE Contract Number: W-7405-ENG-48
• OSTI ID: 5320559
• Report Number(s): NUREG/CR-2189-Vol.3; UCID-18967-Vol.3; ON: TI85016044
• 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
COOLING SYSTEMS
DYNAMIC LOADS
ELASTICITY
ENERGY SYSTEMS
EXPANSION
FAILURES
FINITE ELEMENT METHOD
HIGH TEMPERATURE
JOINTS
LOSS OF COOLANT
MECHANICAL PROPERTIES
MECHANICAL VIBRATIONS
NUMERICAL SOLUTION
PIPES
PRESSURE EFFECTS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
RESIDUAL STRESSES
SEISMIC EFFECTS
STRESS ANALYSIS
STRESSES
TEMPERATURE GRADIENTS
TENSILE PROPERTIES
THERMAL EXPANSION
WATER COOLED REACTORS
WATER MODERATED REACTORS
WELDED JOINTS
YOUNG MODULUS