Compact nuclear reactor with integral steam generator
View Patent ↗In an illustrative embodiment, a pressurized water nuclear reactor (PWR) includes a pressure vessel ( 12, 14, 16 ), a nuclear reactor core ( 10 ) disposed in the pressure vessel, and a vertically oriented hollow central riser ( 36 ) disposed above the nuclear reactor core inside the pressure vessel. A once-through steam generator (OTSG) ( 30 ) disposed in the pressure vessel includes vertical tubes ( 32 ) arranged in an annular volume defined by the central riser and the pressure vessel. The OTSG further includes a fluid flow volume surrounding the vertical tubes and having a feedwater inlet ( 50 ) and a steam outlet ( 52 ). The PWR has an operating state in which feedwater injected into the fluid flow volume at the feedwater inlet is converted to steam by heat emanating from primary coolant flowing inside the tubes of the OTSG, and the steam is discharged from the fluid flow volume at the steam outlet.
1. An apparatus comprising:
a generally cylindrical pressure vessel defining a cylinder axis;
a nuclear reactor core disposed in the generally cylindrical pressure vessel;
a central riser disposed coaxially inside the generally cylindrical pressure vessel, the central riser being hollow and having a bottom end proximate to the nuclear reactor core to receive primary coolant heated by the nuclear reactor core, the central riser having a top end distal from the nuclear reactor core;
a feedwater inlet comprising a vessel penetration of the generally cylindrical pressure vessel;
a steam outlet comprising a vessel penetration of the generally cylindrical pressure vessel; and
a once-through steam generator (OTSG) comprising tubes arranged parallel with the cylinder axis in an annular volume defined between the central riser and the generally cylindrical pressure vessel, primary coolant discharged from the top end of the central riser flowing inside the tubes toward the nuclear reactor core, the OTSG further including a fluid flow volume in fluid communication with the feedwater inlet and with the steam outlet wherein fluid injected into the fluid flow volume at the feedwater inlet and discharged from the fluid flow volume at the steam outlet flows outside the tubes in a direction generally opposite flow of primary coolant inside the tubes;
wherein a feedwater annular space is defined between an outer shroud of the OTSG and the pressure vessel, the fluid flow volume at being in fluid communication with the feedwater inlet via the feedwater annular space; and
wherein a steam annular space is defined between the outer shroud of the OTSG and the pressure vessel, the fluid flow volume at being in fluid communication with the steam outlet via the steam annular space;
wherein the feedwater annular space has a larger outer diameter than the steam annular space.
2. The apparatus as set forth in claim 1 , wherein the outer shroud of the OTSG comprises a cylindrical outer shroud having a constant outer diameter, the feedwater annular space and the steam annular space having the same inner diameter defined by the constant outer diameter of the cylindrical outer shroud of the OTSG.
3. The apparatus as set forth in claim 1 , wherein the apparatus comprises a nuclear reactor having an operating state in which fluid comprising feedwater injected into the fluid flow volume at the feedwater inlet passes through the feedwater annular space into the fluid flow volume where the feedwater is converted by heat transfer from primary coolant flowing inside the tubes into steam that is discharged from the fluid flow volume at the steam outlet after passing through the steam annular space.
4. The apparatus as set forth in claim 3 , wherein in the operating state the OTSG defines an integral economizer that heats feedwater injected into the fluid flow volume at the feedwater inlet to a temperature at or below a boiling point of the feedwater.
5. The apparatus as set forth in claim 3 , further comprising:
a flow diverter disposed in the generally cylindrical pressure vessel, the flow diverter having a flow-diverting surface facing the top end of the central riser that is at least one of sloped and curved to redirect primary coolant discharged from the top end of the central riser toward inlets of the tubes of the OTSG.
6. The apparatus as set forth in claim 3 , further comprising:
A perforated cylinder flow diverter disposed in the generally cylindrical pressure vessel, wherein the perforated cylinder flow diverter extends vertically from an upper portion of the central riser.
7. The apparatus as set forth in claim 3 , further comprising:
a divider plate spaced apart from the top end of the central riser;
wherein the generally cylindrical pressure vessel includes a sealing top portion that together with the divider plate defines an integral pressurizer volume that is separated by the divider plate from the remaining interior volume of the generally cylindrical pressure vessel; and
wherein in the operating state of the nuclear reactor comprises a pressurized water reactor (PWR) and the integral pressurizer volume contains fluid at a temperature greater than a temperature of the primary coolant disposed in the remaining interior volume of the generally cylindrical pressure vessel.
8. The apparatus as set forth in claim 7 , wherein the divider plate comprises:
a flow diverter having a flow-diverting surface facing the top end of the central riser that is at least one of sloped and curved to redirect primary coolant discharged from the top end of the central riser toward inlets of the tubes of the OTSG.
9. The apparatus as set forth in claim 7 , further comprising:
neutron-absorbing control rods; and
a control rod drive mechanism (CRDM) configured to controllably insert and withdraw the control rods into and out of the nuclear reactor core;
wherein no portion of the CRDM is disposed in or passes though the integral pressurizer volume.
10. The apparatus as set forth in claim 3 , wherein:
in the operating state of the nuclear reactor the primary coolant flowing inside the tubes is at a higher pressure than the fluid in the fluid flow volume.
11. The apparatus as set forth in claim 3 , wherein:
in the operating state of the nuclear reactor the primary coolant flowing inside the tubes is at a pressure that is at least twice a pressure of the fluid in the fluid flow volume.
12. An apparatus comprising:
a pressurized water nuclear reactor (PWR) including a pressure vessel, a nuclear reactor core disposed in the pressure vessel, and a vertically oriented hollow central riser disposed above the nuclear reactor core inside the pressure vessel;
a once-through steam generator (OTSG) disposed in the pressure vessel of the PWR, the OTSG including vertical tubes arranged in an annular volume defined between the central riser and the pressure vessel, the OTSG further including a fluid flow volume surrounding the vertical tubes, the OTSG further including an outer shroud surrounding vertical tubes of the OTSG arranged in the annular volume defined between the central riser and the pressure vessel;
a feedwater inlet comprising a vessel penetration of the pressure vessel; and
a steam outlet comprising a vessel penetration of the pressure vessel;
wherein a feedwater annular space is defined between the outer shroud of the OTSG and the pressure vessel, and feedwater injected at the feedwater inlet passes through the feedwater annular space into the fluid flow volume, and
wherein a steam annular space is defined between the outer shroud of the OTSG and the pressure vessel, and steam discharged from the fluid flow volume passes through the steam annular space and exits at the steam outlet.
13. The apparatus as set forth in claim 12 wherein the feedwater annular space has a larger outer diameter than the steam annular space.