Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

Conference · Thu Jan 01 04:00:00 EST 1981

OSTI ID:5278951

Lu, S C; Larder, R A; Chan, A L; Ma, M S

This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station (Zion-1). The results of the analyses were used as input to a simulation procedure for predicting pipe rupture probabilities of the reactor coolant system presented by another SMiRT-6 paper J6/6. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, the reactor pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations.

• Research Organization: Lawrence Livermore National Lab., CA (USA); URS/John A. Blume and Associates, Engineers, San Francisco, CA (USA); Bechtel Corp., San Francisco, CA (USA)
• DOE Contract Number: W-7405-ENG-48
• OSTI ID: 5278951
• Report Number(s): UCRL-84190; CONF-810801-24; ON: TI85016700
• 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
COOLING SYSTEMS
ENERGY SYSTEMS
ENRICHED URANIUM REACTORS
FAILURES
PIPES
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PROBABILITY
PWR TYPE REACTORS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
RUPTURES
STRESS ANALYSIS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZION-1 REACTOR