Nuclear reactor system having natural circulation of primary coolant

Abstract

A nuclear reactor system that, in one embodiment, utilizes natural circulation (i.e., thermosiphon) to circulate a primary coolant in a single-phase through a reactor core and a heat exchange sub-system. The heal exchange sub-system is located outside of the nuclear reactor pressure vessels and, in some embodiments, is designed so as to not cause any substantial pressure drop in the flow of the primary coolant within the heal exchange sub-system that is used to vaporize a secondary coolant. In another embodiment, a nuclear reactor system is disclosed in which the reactor core is located below ground and all penetrations into the reactor pressure vessel are located above ground.

1 . A natural circulation nuclear reactor system comprising:
 a reactor pressure vessel having an internal cavity;   a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel;   a heat exchange sub-system located outside of the reactor pressure vessel;   a closed-loop primary coolant circuit that flows a primary coolant through the reactor pressure vessel to cool the reactor core and through the heat exchange sub-system to transfer heat to a secondary coolant; and   wherein operation of the reactor core causes natural circulation of the primary coolant through the closed-loop primary coolant circuit in a single phase.   
     
     
         2 . The natural circulation nuclear reactor system according to  claim 1  further comprising:
 a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway, wherein operation of the reactor core causes the primary coolant to rise within the primary coolant riser passageway; 
 the reactor pressure vessel comprising a primary coolant outlet port located at a top portion of the reactor pressure vessel in fluid Communication with the primary coolant riser passageway; 
 the reactor pressure vessel comprising a primary coolant inlet port located at a top portion of the reactor pressure vessel in fluid communication with the primary coolant downcomer passageway; and 
 a plenum at the bottom portion of the reactor vessel that allows cross-flow of the primary coolant from the primary coolant downcomer passageway to the primary coolant riser passageway. 
 
     
     
         3 . (canceled) 
     
     
         4 . The natural circulation nuclear reactor system according to  claim 2  wherein the partition has an effective coefficient of thermal conductivity measured from the primary coolant riser passageway to the primary coolant downcomer passageway that is less than an effective coefficient of thermal conductivity of the primary coolant. 
     
     
         5 . The natural circulation nuclear reactor system according to  claim 2  wherein the partition is a tubular structure having an inner surface and an outer surface, the inner surface of the tubular structure forming the primary coolant riser passageway, and the primary coolant downcomer passageway being formed between the outer surface of tubular structure and the inner surface of the reactor pressure vessel, the primary coolant downcomer passageway being an annular passageway that circumferentially surrounds the primary coolant riser passageway. 
     
     
         6 .- 8 . (canceled) 
     
     
         9 . The natural circulation nuclear reactor system according to  claim 1  wherein the reactor pressure vessel extends along a substantially vertical axis, a major portion of the axial length of the reactor pressure vessel located below a ground level, and the reactor core located below the ground level and the heat exchange sub-system located above the ground level. 
     
     
         10 . The natural circulation nuclear reactor system according to  claim 9  further comprising:
 a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway; 
 the reactor pressure vessel comprising a primary coolant outlet port located at a top portion of the reactor pressure vessel above the ground level and in fluid communication with the primary coolant riser passageway, the primary coolant outlet port fluidly coupled to the heat exchange sub-system to form an incoming hot leg of primary coolant of the heat exchange sub-system; and 
 the reactor pressure vessel comprising a primary coolant inlet port located at the top portion of the reactor pressure vessel above the ground level and in fluid communication with the primary coolant downcomer passageway, the primary coolant inlet port fluidly coupled to the heat exchange sub-system to form an outgoing cool leg of primary coolant of the heat exchange sub-system. 
 
     
     
         11 . The natural circulation nuclear reactor system according to  claim 10  wherein the primary coolant has a first high temperature in the incoming hot lee and a second low temperature in the outgoing cool leg, the first high temperature being at least 220° F. greater than the second low temperature. 
     
     
         12 .- 14 . (canceled) 
     
     
         15 . The natural circulation nuclear reactor system according to  claim 9  wherein the reactor pressure vessel comprises a flange portion that supports the major portion of the reactor pressure vessel below the ground level in a vertically-oriented cantilevered manner within a reactor well. 
     
     
         16 .- 19 . (canceled) 
     
     
         20 . The natural circulation nuclear reactor system according to  claim 1  wherein the heat exchange sub-system comprises a first horizontal steam generator and a second horizontal steam generator operably coupled in series with one another other along the closed-loop primary coolant circuit, the first and second steam generators converting the secondary coolant from liquid-phase to gas-phase via the heat transferred from the primary coolant. 
     
     
         21 . The natural circulation nuclear reactor system according to  claim 20  wherein the first horizontal steam generator is a high pressure steam generator and the second horizontal steam generator is a low pressure steam generator, the high pressure steam generator being upstream of the low pressure steam generator along the closed-loop primary coolant circuit. 
     
     
         22 . (canceled) 
     
     
         23 . The natural circulation nuclear reactor system according to  claim 20  wherein the first and second horizontal steam generators are single-pass heat exchangers, the primary coolant being a tube-side fluid in both the first and second horizontal steam generators, and wherein the first and second horizontal steam generators do not cause any substantial pressure drop in the closed-loop primary coolant circuit due to increase in elevation. 
     
     
         24 . (canceled) 
     
     
         25 . The natural circulation nuclear reactor system according to  claim 20  wherein an inlet of the first horizontal steam generator is coupled directly to a primary coolant outlet port of the reactor pressure vessel, an outlet of the first horizontal steam generator is coupled directly to an inlet of the second horizontal steam generator, and an outlet of the second horizontal steam generator is coupled directly to a primary coolant inlet port of the reactor pressure vessel. 
     
     
         26 . (canceled) 
     
     
         27 . The natural circulation nuclear reactor system according to  claim 25  wherein the first and second horizontal steam generators are integrally welded to the reactor vessel. 
     
     
         28 . The natural circulation nuclear reactor system according to  claim 20  wherein the heat exchange sub-system comprises a first superheater configured to heat the vapor-phase of the secondary coolant exiting the first steam generator and a second superheater configured to heat the vapor-phase of the secondary coolant exiting the second steam generator, wherein the first and second superheaters are operably coupled to one another in series along the closed-loop primary coolant circuit and in parallel to the first and second steam generators along the closed-loop primary coolant circuit. 
     
     
         29 . The natural circulation nuclear reactor system according to  claim 28  wherein between 10% to 15% of the flow of the primary coolant through the heat exchange sub-system is directed through the first and second superheaters, the remainder of the flow of the primary coolant through the heat exchange sub-system being directed through the first and second steam generators. 
     
     
         30 . (canceled) 
     
     
         31 . The natural circulation nuclear reactor system according to  claim 1  wherein operation of the reactor core causes natural circulation of the primary coolant by creating a riser water column and a downcomer water column within the internal cavity of the reactor pressure vessel, the riser water column and the downcomer water column having a vertical height in a range of 80 ft. to 120 ft. 
     
     
         32 . The natural circulation nuclear reactor system according to  claim 1  wherein the primary coolant has a negative reactivity coefficient. 
     
     
         33 . (canceled) 
     
     
         34 . (canceled) 
     
     
         35 . The natural circulation nuclear reactor system according to  claim 1  wherein the internal cavity of the reactor pressure vessel is maintained at a pressure in a range of 2000 psia to 2500 psia. 
     
     
         36 . (canceled) 
     
     
         37 . The natural circulation nuclear reactor system according to  claim 1  wherein the heat exchange sub-system is at an elevation that is 80 ft. to 120 ft. greater than an elevation of the reactor core. 
     
     
         38 . (canceled) 
     
     
         39 . A nuclear reactor system comprising:
 an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis, a major portion of the axial length of the reactor pressure vessel located below a ground level;   a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor and below the ground level;   the reactor pressure vessel comprising a primary coolant outlet port located above the ground level;   the reactor pressure vessel comprising a primary coolant inlet port located above the ground level;   a heat exchange sub-system located outside of the reactor pressure vessel and above the ground level, an incoming hot leg of the heat exchange system fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the heat exchange system fluidly coupled to the primary coolant inlet port; and   wherein the major portion of the reactor pressure vessel is free of penetrations.   
     
     
         40 . The nuclear reactor system according to  claim 39  further comprising:
 a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway; 
 the primary coolant outlet port in fluid communication with a top portion of the primary coolant riser passageway ; and 
 the primary coolant inlet port in fluid communication with a top portion of the primary coolant downcomer passageway. 
 
     
     
         41 . The nuclear reactor system according to  claim 40  wherein the partition has an effective coefficient of thermal conductivity measured from the primary coolant riser passageway to the primary coolant downcomer passageway that is less than an effective coefficient of thermal conductivity of the primary coolant. 
     
     
         42 . The nuclear reactor system according to  claim 40  wherein the primary coolant has a first high temperature in the incoming hot leg and a second low temperature in the outgoing cool leg, the first high temperature being at least 220° F. greater than the second low temperature. 
     
     
         43 . The nuclear reactor system according to  claim 39  wherein the major portion of the reactor pressure vessel is at least 75% of the axial length of the reactor pressure vessel. 
     
     
         44 . (canceled) 
     
     
         45 . (canceled) 
     
     
         46 . The nuclear reactor system according to  claim 39  wherein the heat exchange sub-system and the internal cavity of the reactor pressure vessel collectively form a closed-loop primary coolant circuit that flows the primary coolant through the reactor pressure vessel to cool the reactor core and through the heat exchange sub-system to transfer heat to a secondary coolant, wherein operation of the reactor core causes natural circulation of the primary coolant through the closed-loop primary coolant circuit in a single phase, and wherein the secondary coolant is converted from liquid-phase to gas-phase within the heat exchange sub-system by the heat transferred from the primary coolant. 
     
     
         47 . (canceled) 
     
     
         48 . The nuclear reactor system according to  claim 46  wherein the heat exchange sub-system comprises a first horizontal steam generator and a second horizontal steam generator operably coupled in series with one another other along the closed-loop primary coolant circuit, the first and second steam generators converting the secondary coolant from liquid-phase to gas-phase via the heat transferred from the primary coolant. 
     
     
         49 . The nuclear reactor system according to  claim 48  wherein the first and second horizontal steam generators are single-pass heat exchangers, the primary coolant being a tube-side fluid in both the first and second horizontal steam generators, and wherein the first and second horizontal steam generators do not cause any substantial pressure drop in the closed-loop primary coolant circuit due to increase in elevation. 
     
     
         50 . The nuclear reactor system according to  claim 48  wherein an inlet of the first horizontal steam generator is coupled directly to the primary coolant outlet port of the reactor pressure vessel, an outlet of the first horizontal steam generator is coupled directly to an inlet of the second horizontal steam generator, and an outlet of the second horizontal steam generator is coupled directly to the primary coolant inlet port of the reactor pressure vessel. 
     
     
         51 . (canceled) 
     
     
         52 . (canceled) 
     
     
         53 . The nuclear reactor system according to  claim 46  wherein the primary coolant has a negative reactivity coefficient. 
     
     
         54 . A nuclear reactor system comprising:
 an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis;   a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor;   a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway;   the reactor pressure vessel comprising a primary coolant outlet port in fluid communication with a top portion of the primary coolant riser passageway;   the reactor pressure vessel comprising a primary coolant inlet port in fluid communication with a top portion of the primary downcomer riser passageway;   at least one steam generator located outside of the reactor pressure vessel, an incoming hot leg of the steam generator fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the steam generator fluidly coupled to the primary coolant inlet port; and   wherein the steam generator does not cause any substantial pressure drop in a flow of the primary coolant through the steam generator resulting from an increase in elevation.   
     
     
         55 . (canceled) 
     
     
         56 . (canceled)