Stress analysis, code evaluation, and design modification of a plate resulting from seismic loads and hypothetical core disruptive accident loads
This study addresses the structural analysis and evaluation of a design modification for a plate in the Fast Flux Test Facility heat temperature control system. The plate being considered is near a fuel transfer port system. The plate is flat and is supported by six long studs, five of which are along one side of the plate. Their location makes the plate act as a cantilever.The plate itself provides support to three vertical neutron shields on its free edges. During service, a uranium shield ring under the fuel transfer port nozzle oxidized and expanded. To prevent this expansion from causing damage to the surrounding components, this ring was removed and replaced with lead blocks. Approximately one-fourth of the lead blocks rest on a free edge of the plate. This new configuration of the plate required updated seismic and hypothetical core disruptive accident analyses. Seismic and hypothetical core disruptive accident analyses were performed and checked against the requirements of the American Society of Mechanical Engineers Code. The result showed that the design of the existing plate supports (the six studs) was not adequate for the added weight of the lead blocks. The design was modified to restrain the lead blocks. When the final design with modified boundary conditions was reevaluated, the stress results satisfied the Code requirements.
- Research Organization:
- Westinghouse Hanford Co., Richland, WA (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC06-87RL10930
- OSTI ID:
- 5667247
- Report Number(s):
- WHC-SA-1337; CONF-920631-11; ON: DE92008363
- Resource Relation:
- Conference: American Society of Mechanical Engineers pressure vessel and piping conference, New Orleans, LA (United States), 21-25 Jun 1992
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
FUEL PLATES
STRESS ANALYSIS
ACCIDENTS
DESIGN
ENGINEERING DRAWINGS
FFTF REACTOR
MATHEMATICAL MODELS
REACTOR CORE DISRUPTION
REACTOR SAFETY
REGULATIONS
RISK ASSESSMENT
SEISMIC EFFECTS
SHIELD SUPPORTS
STANDARDS
DIAGRAMS
EPITHERMAL REACTORS
FAST REACTORS
FUEL ELEMENTS
LIQUID METAL COOLED REACTORS
MECHANICAL STRUCTURES
POWERED SUPPORTS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
SUPPORTS
TEST REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220900 - Nuclear Reactor Technology- Reactor Safety