Piping support and gate valve behavior during high level HDR (Heissdampfreaktor) simulated seismic tests
This paper describes a portion of the analysis and results to date from the SHAM seismic research program, in which the USNRC/INEL was a participant. The program was conducted by Kernforschungszentrum Karlsruhe (KfK) at the decommissioned Heissdampfreaktor (HDR) located near Frankfurt in the Federal Republic of Germany (FRG). The research program included the study of the effects of increasing levels of seismic excitation on a full scale, in situ nuclear piping system containing a naturally aged US 8-in. motor-operated gate valve. In all, 51 experiments were conducted with the piping supported by six different piping support systems, including a typical stiff US piping support system made up of snubbers and rigid struts. Earthquake-like displacement histories were input to servohydraulic shakers attached directly to the piping system. Inputs to the piping system started with a magnitude of 0.6 g ZPA and were stepped up, using the same frequency content, to a maximum of 5 g ZPA. The resulting piping system response produced overload snubber failures and local piping strains of 0.9%. The results of this testing will contribute to the technical basis used for support and development of equipment qualification standards and will help establish the seismic safety margins in piping and piping support system components. 13 figs., 3 tabs.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6796818
- Report Number(s):
- EGG-M-88295; CONF-8810155-37; ON: DE89005410
- Country of Publication:
- United States
- Language:
- English
Similar Records
Piping system response during high-level simulated seismic tests at the Heissdampfreaktor Facility: (SHAM Test Facility)
Piping system response during high-level simulated seismic tests at the Heissdampfreaktor Facility: (SHAM Test Facility)
Related Subjects
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
BWR TYPE REACTORS
CONTAINERS
CONTROL EQUIPMENT
DYNAMIC LOADS
ENRICHED URANIUM REACTORS
EQUIPMENT
EUROPE
EXPERIMENTAL REACTORS
FASTENERS
FEDERAL REPUBLIC OF GERMANY
FLOW REGULATORS
HDR REACTOR
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
PIPES
POWER PLANTS
PRESSURE VESSELS
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH PROGRAMS
RESTRAINTS
SAFETY
SEISMIC EFFECTS
STANDARDS
TESTING
THERMAL POWER PLANTS
THERMAL REACTORS
VALVES
WATER COOLED REACTORS
WATER MODERATED REACTORS
WESTERN EUROPE