Thermohydraulics in a high temperature gas-cooled reactor prestressed concrete reactor vessel during unrestricted core heatup accidents
Copyright policy )Thermohydraulics in a high temperature gas-cooled reactor prestressed concrete reactor vessel during unrestricted core heatup accidents
Abstract The long term core and primary loop heatup of an HTGR subsequent to loss of all forced circulation has been analyzed using a modified version of the CORCON code. The results indicate that if the liner cooling system is operating, or can be restarted within about 60 h, safe cooldown can be achieved, but significant core damage will occur. Without functioning liner cooling system the core heatup will lead to PCRV concrete degradation and the resulting concrete gas releases will ultimately cause containment building failure after 6 to 10 days.
- University of North Texas United States
- Brookhaven National Laboratory United States
- Brookhaven National Laboratory United States
- University of North Texas United States
Nonbreeding, Temperature Gradients, Shields, Hydraulics, Fluid Mechanics, 21 Specific Nuclear Reactors And Associated Plants, Mechanics, Reactor Accidents, Pressure Vessels, Containers, Gas Cooled Reactors, 210300 -- Power Reactors, Concretes, Building Materials, Loss Of Flow, Thermal Shields, Materials, 22 General Studies Of Nuclear Reactors, Reactor Safety, After-Heat, Chemical Reactions, Heat Transfer, Reactors, Thermal Stresses, Energy Transfer, Accidents, Graphite Moderated Reactors, Graphite Moderated, Stresses 220900* -- Nuclear Reactor Technology-- Reactor Safety, Safety, Fuel Element Failure, Htgr Type Reactors
Nonbreeding, Temperature Gradients, Shields, Hydraulics, Fluid Mechanics, 21 Specific Nuclear Reactors And Associated Plants, Mechanics, Reactor Accidents, Pressure Vessels, Containers, Gas Cooled Reactors, 210300 -- Power Reactors, Concretes, Building Materials, Loss Of Flow, Thermal Shields, Materials, 22 General Studies Of Nuclear Reactors, Reactor Safety, After-Heat, Chemical Reactions, Heat Transfer, Reactors, Thermal Stresses, Energy Transfer, Accidents, Graphite Moderated Reactors, Graphite Moderated, Stresses 220900* -- Nuclear Reactor Technology-- Reactor Safety, Safety, Fuel Element Failure, Htgr Type Reactors
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