Compact nuclear reactor with integral steam generator

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 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 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.

A nuclear reactor 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;
• 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 having a feedwater inlet and a 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 the nuclear reactor has an operating state in which fluid comprising feedwater injected into the fluid flow volume at the feedwater inlet 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;
• and
• wherein the tubes of the OTSG are secured in a support made of steel and supported at their ends by tubesheets;
• wherein the tubes comprise a material having a higher coefficient of thermal expansion than steel and are in axial tension in a non-operating state of the nuclear reactor in which the tubes of the OTSG are at a temperature of less than 100° C. and are in axial compression in the operating state.

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.

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;
• and
• a once-through steam generator (OTSG) disposed in the pressure vessel of the PWR, the OTSG including vertical tubes having a higher coefficient of thermal expansion than steel arranged in an annular volume defined by the central riser and the pressure vessel and secured in a support made of steel, the OTSG further including a fluid flow volume surrounding the vertical tubes;
• wherein the PWR has an operating state in which primary coolant at a temperature of at least 500° C. flows in the tubes of the OTSG and in which feedwater injected into the fluid flow volume at a feedwater inlet is converted to steam by heat emanating from the primary coolant flowing inside the tubes of the OTSG, and the steam is discharged from the fluid flow volume at a steam outlet;
• and
• wherein the tubes of the OTSG are in axial tension in a non-operating state of the PWR in which the tubes of the OTSG are at a temperature of less than 100° C. and are in axial compression in the operating state.