Abstract: Arrangement for supporting U-bend tube sections in the high heat environment of steam generators using flat bars. The invention uses a combination of thicker and thinner flat bars to impart a serpentine path to the arc of the normally curvilinear U-tubes. The support system accommodates the dilation and contraction of coolant tubes and other elements caused by the extreme and varying conditions inside a steam generator, and which can cause gaps between coolant tubes and prior art tube support bars. Bars of alternating thickness provide alternating offsets to tensionally push and support each tube on multiple sides and in multiple locations, and this tension keeps the tubes in contact with at least some flat bars on multiple sides regardless of size and shape changes. Support arrangement includes a set of fan bars, each fan bar including thick and thin flat bars projecting up and out from a collector bar.

Abstract: Apparatus for a steam generator that employs tube support plates within a shroud that is in turn disposed within a shell. The tube support plates are made of a material having a coefficient of thermal expansion lower than that of the shroud. The tube support plates are aligned during fabrication, with minimal clearances between components. Using a tube support displacement system, a controlled misalignment is then imposed on one or more tube support plates, as the steam generator heats up. The tube support plate displacement system has only one part, a push rod, which is internal to the steam generator shroud, thereby minimizing the potential of loose parts. The tube support plate displacement system can be used to provide controlled misalignments on one or more tube support plates, in the same or varying amounts and directions, and with one or more apparatus for each individual tube support plate.

Abstract: Arrangement for supporting U-bend tube sections in the high heat environment of steam generators using flat bars. The invention uses a combination of thicker and thinner flat bars to impart a serpentine path to the arc of the normally curvilinear U-tubes. The support system accommodates the dilation and contraction of coolant tubes and other elements caused by the extreme and varying conditions inside a steam generator, and which can cause gaps between coolant tubes and prior art tube support bars. Bars of alternating thickness provide alternating offsets to tensionally push and support each tube on multiple sides and in multiple locations, and this tension keeps the tubes in contact with at least some flat bars on multiple sides regardless of size and shape changes. Support arrangement includes a set of fan bars, each fan bar including thick and thin flat bars projecting up and out from a collector bar.

Abstract: 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.

Abstract: A component, such as a cyclonic steam separator baseplate of a steam generator, includes a surface subject to degradation during operation of the system in which the component is disposed. A profile is acquired of the surface of the component using an optical surface profilometry system concurrent with an image of the surface. A condition, such as degradation of the component is classified based on the acquired profile and image of the surface of the component. Component conditions may be monitored over time, trended, and classified as requiring maintenance, repair, or replacement.

Abstract: A pressurized water reactor (PWR) comprises: a nuclear core comprising a fissile material; a cylindrical pressure vessel having a vertically oriented cylinder axis and containing the nuclear core immersed in primary coolant water; and a hollow cylindrical central riser disposed concentrically with and inside the cylindrical pressure vessel. A downcomer annulus is defined between the hollow cylindrical central riser and the cylindrical pressure vessel. The hollow cylindrical central riser has a radially expanding upper orifice that merges into an annular divider plate that separates an upper plenum above the annular divider plate from a lower plenum below the annular divider plate. The upper plenum is in fluid communication with the radially expanding upper orifice and the lower plenum is in fluid communication with the downcomer annulus. A weir may extend away from a bottom wall of the lower plenum into the lower plenum.

Abstract: Apparatus for a steam generator that employs tube support plates within a shroud that is in turn disposed within a shell. The tube support plates are made of a material having a coefficient of thermal expansion lower than that of the shroud. The tube support plates are aligned during fabrication, with minimal clearances between components. Using a tube support displacement system, a controlled misalignment is then imposed on one or more tube support plates, as the steam generator heats up. The tube support plate displacement system has only one part, a push rod, which is internal to the steam generator shroud, thereby minimizing the potential of loose parts. The tube support plate displacement system can be used to provide controlled misalignments on one or more tube support plates, in the same or varying amounts and directions, and with one or more apparatus for each individual tube support plate.

Abstract: An apparatus in a steam generator that employs tube support plates within a shroud that is in turn disposed within a shell. The tube support plates are made of a material having a coefficient of thermal expansion lower than that of the shroud. The tube support plates are aligned during fabrication with minimal clearances between components. Using a tube support displacement system, a controlled misalignment is imposed on one or more tube support plates as the steam generator heats up. The tube support plate displacement system has only two parts, a spring bar and a push rod, that are internal to the steam generator shell and threadably engaged, thereby minimizing the potential of loose parts. The tube support plate displacement system can be used to provide controlled misalignments on one or more tube support plates, with one or more apparatus being provided for any individual tube support plate.

Abstract: 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.

Abstract: 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.