Simulation experiments for a large leak sodium-water reaction analysis. Volume 4. IHTS/relief system simulation tests

Technical Report · Fri Sep 01 00:00:00 EDT 1978

DOI:https://doi.org/10.2172/6031355· OSTI ID:6031355

Ploeger, D W

Tests were performed in which a simplified 1/8-scale model of the intermediate heat transfer system and relief system of a LMFBR was subjected to a simulated sodium-water reaction in a steam generator. Pressures in the intermediate heat exchanger (IHX) and in the pipe were measured. The flow of water through the relief system was photographed and its velocity was measured. The forces on the relief system elbows resulting from the fluid flow were also measured. The tests were performed primarily to validate pulse propagation codes used for design and for direct use as design data.

• Research Organization: Stanford Research Inst., Menlo Park, CA (USA)
• DOE Contract Number: AM03-76SF00115
• OSTI ID: 6031355
• Report Number(s): SRI-PYD-6272-Vol.4; ON: DE82004237; TRN: 82-000610
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LMFBR TYPE REACTORS
MOLTEN METAL-WATER REACTIONS
REACTOR SAFETY EXPERIMENTS
FLUID FLOW
HEAT EXCHANGERS
LEAKS
MOCKUP
PRESSURIZING
REACTOR COOLING SYSTEMS
RELIEF VALVES
SIMULATION
SODIUM
STEAM GENERATORS
VELOCITY
WATER
ALKALI METALS
BOILERS
BREEDER REACTORS
CONTROL EQUIPMENT
COOLING SYSTEMS
ELEMENTS
ENERGY SYSTEMS
EPITHERMAL REACTORS
EQUIPMENT
FAST REACTORS
FBR TYPE REACTORS
FLOW REGULATORS
HYDROGEN COMPOUNDS
LIQUID METAL COOLED REACTORS
METALS
OXYGEN COMPOUNDS
REACTOR COMPONENTS
REACTORS
STRUCTURAL MODELS
VALVES
VAPOR GENERATORS
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
Breeding