Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 2: primary coolant loop model. Final report

Technical Report · Tue Sep 01 00:00:00 EDT 1981

OSTI ID:5590145

Eberhardt, A C

This report describes the Zion Station reactor coolant loop model developed by Sargent and Lundy Engineers for Lawrence Livermore National Laboratory as part of its Load Combination Program. This model was developed for use in performing seismic time history analyses of an actual pressurized water reactor (PWR) system. It includes all major items affecting the seismic response of a 4-loop Westinghouse nuclear steam supply system: the components, supports, and interconnecting piping. The model was further expanded to permit static analysis of dead weight, thermal, and internal pressure load conditions.

• Research Organization: Lawrence Livermore National Lab., CA (USA)
• OSTI ID: 5590145
• Report Number(s): NUREG/CR-2189-V2; TRN: 82-008597
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PRIMARY COOLANT CIRCUITS
SEISMIC EFFECTS
PWR TYPE REACTORS
COMPUTERIZED SIMULATION
DYNAMIC LOADS
EARTHQUAKES
MATHEMATICAL MODELS
PIPES
PROBABILITY
RUPTURES
COOLING SYSTEMS
ENERGY SYSTEMS
FAILURES
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SEISMIC EVENTS
SIMULATION
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled