Verification of piping response calculations of SMACS code with data from seismic testing of an in-plant piping system

Technical Report · Sun Sep 01 04:00:00 EDT 1991

DOI:https://doi.org/10.2172/5244469· OSTI ID:5244469

Srinivasan, M G; Kot, C A; Hsieh, B J ^[1]

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1. Argonne National Lab., IL (United States)

The objective of this effort was to evaluate the piping analysis part of the SMACS code for estimating the response of realistic piping systems subjected to seismic excitation, given as multiple, independent support acceleration histories. The experimental data from the seismic testing of an in-plant piping system at the HDR Test Facility in Germany were used for this purpose. Of the six different support systems tested, two were selected for the evaluation: one a stiff'' configuration containing both struts and snubbers, and the other, a more flexible configuration with no snubbers. Described are the analytical modeling, calculations, and results of the posttest simulation of two sets each for both support configurations, with excitations at 100% and 200/300% of safe-shutdown-earthquake loading. Almost all the calculated peak response quantities were smaller (by different amounts) than the corresponding test measurements. However, pipe displacements and bending stresses were better estimated than the pipe accelerations and support forces. The discrepancies are mainly attributable to the inability of the linear analysis to model the nonlinear behavior of the VKL piping system, characterized by gaps in support connections and the friction at pipe clamps. A similar trend of underestimating test responses was observed in the linear analyses performed by other investigators. 13 refs., 14 figs., 8 tabs.

• Research Organization: Argonne National Lab., IL (United States); Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering
• Sponsoring Organization: NRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 5244469
• Report Number(s): NUREG/CR-5757; ANL--91/25; ON: TI91018982
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCELERATION
BENDING
COMPUTER CALCULATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
DAMPING
DATA
DEFORMATION
DYNAMIC LOADS
EXPERIMENTAL DATA
FUNCTIONS
INFORMATION
MECHANICAL STRUCTURES
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
NUMERICAL DATA
PIPES
POWER PLANTS
RESPONSE FUNCTIONS
RESTRAINTS
S CODES
SEISMIC EFFECTS
SIMULATION
STRAINS
STRESS ANALYSIS
SUPPORTS
THERMAL POWER PLANTS