Summary of thermal-hydraulic calculations for a pressurized water reactor
Summary of thermal-hydraulic calculations for a pressurized water reactor
The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid. Calculations show that continued injection of high pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island.
- University of North Texas United States
Loss Of Coolant, Eccs, Nonbreeding, Nonboiling Water Cooled, Equipment, 21 Specific Nuclear Reactors And Associated Plants, Reactor Accidents, Transients, Energy Losses, Relief Valves, Light-Water Moderated, High Pressure Coolant Injection, 22 General Studies Of Nuclear Reactors, Losses, Computer Codes, Power Losses, Pwr Type Reactors, Reactors, Valves, Steam Generators, Reactor Protection Systems, Control Equipment, Vapor Generators, Flow Regulators, Engineered Safety Systems, Steam, Accidents, Bubbles, 210200 -- Power Reactors, Water Cooled Reactors, Boilers, Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety
Loss Of Coolant, Eccs, Nonbreeding, Nonboiling Water Cooled, Equipment, 21 Specific Nuclear Reactors And Associated Plants, Reactor Accidents, Transients, Energy Losses, Relief Valves, Light-Water Moderated, High Pressure Coolant Injection, 22 General Studies Of Nuclear Reactors, Losses, Computer Codes, Power Losses, Pwr Type Reactors, Reactors, Valves, Steam Generators, Reactor Protection Systems, Control Equipment, Vapor Generators, Flow Regulators, Engineered Safety Systems, Steam, Accidents, Bubbles, 210200 -- Power Reactors, Water Cooled Reactors, Boilers, Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety
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