Heat-extraction system for gas-cooled nuclear reactor

A gas-cooled nuclear reactor including

• a reactor core,
• a pressure vessel in which said core is located,
• heat-exchange means in said pressure vessel,
• primary circulator means for circulating gas coolant through said reactor core and then through said heat-exchange means, and
• means including secondary circulator means for extracting heat from said reactor by causing a secondary gaseous coolant stream to flow through said heat-exchange means,
• wherein the improvement comprises
• said heat-exchange means raises said secondary gaseous coolant to a temperature suitable to supply heat for process heat applications,
• said primary circulator means being arranged to discharge the repressurized primary gas coolant along an inner boundary region of said pressure vessel,
• said pressure vessel having a cavity which is divided into upper and lower chambers by a horizontally-extending intermediate floor which supports said core thereabove and with said heat-exchange means and primary circuit circulator means being disposed in said lower chamber,
• said pressure vessel being circular in horizontal cross section and said floor being circular and of smaller diameter to provide an annular passageway which interconnects said upper and lower chambers and through which passageway the primary coolant discharge from said primary circulator means is returned to said reactor core,
• said heat extraction means including a plurality of secondary gas coolant circuits connected to said heat-exchange means, and
• emergency cooling means associated with each of said secondary circuits including heat dump means exterior of said pressure vessel for promptly providing relatively low temperature, secondary gas coolant in said circuit for emergency heat removal from said core whereby said secondary gas circulator means in addition to extracting the process heat from said reactor during normal operations also functions to extract heat from said reactor during emergency cooling.