A thermal control system for a reactor pressure vessel comprises a plate having a substantially circular shape that is attached to a wall of the reactor pressure vessel. The plate divides the reactor pressure vessel into an upper reactor pressure vessel region and a lower reactor pressure vessel region. Additionally, the plate is configured to provide a thermal barrier between a pressurized volume located within the upper reactor pressure vessel region and primary coolant located within the lower reactor pressure vessel region. One or more plenums provide a passageway for a plurality of heat transfer tubes to pass through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the plate.

The present invention provides an air circulating sleeve device that is provided below a steam generator to prevent thermal deformation of a sliding base that supports the steam generator of a nuclear reactor, the air circulating sleeve device comprising: a through-hole formed at the center of the sliding base; and a sleeve vertically mounted so as to be aligned with the through-hole, wherein thermal deformation of the sliding base is prevented by performing natural cooling by introducing external air below the sliding base into a stagnated air area inside the sliding base and a skirt support through the sleeve, and the skirt support includes at least one vent hole such that the stagnated air area inside the sliding base and the skirt support is exposed to the air outside the skirt support, and natural circulation of air is performed through the vent hole.

A Ni—Cr—Fe alloy with improved resistance to stress corrosion cracking in nuclear environments, the alloy comprising 23-28 wt % Cr, 25-35 wt % Ni, <0.03 wt % C, <0.70 wt % Si, <1.0 wt % Mn, <0.015 wt % S, >0.35 wt % Ti, 0.15-0.45 wt % Al, <0.75 wt % Cu, and balance Fe and incidental impurities. The alloy may be used in steam generator tubing of a nuclear reactor. A method of producing an article includes: providing the alloy as disclosed herein; forming the alloy into the article by cold working the alloy to 20%; and heat treating the article.

Combined cleanup and heat sink systems work with nuclear reactor coolant loops. Combined systems may join hotter and colder sections of the coolant loops in parallel with any steam generator or other extractor and provide optional heat removal between the same. Combined systems also remove impurities or debris from a fluid coolant without significant heat loss from the coolant. A cooler in the combined system may increase in capacity or be augmented in number to move between purifying cooling and major heat removal from the coolant, potentially as an emergency cooler. The cooler may be joined to the hotter and colder sections through valved flow paths depending on desired functionality. Sections of the coolant loops may be fully above the cooler, which may be above the reactor, to drive flow by gravity and enhance isolation of sections of the coolant loop.

A heat exchanger includes a body having an inlet header through which a fluid is introduced, and an outlet header through which the fluid is discharged; and one or more plates accommodated in the body and provided with flow path modules providing flow paths for the fluid introduced through the inlet header to flow to the outlet header. The heat exchanger further includes at least one flow path adjuster each having at least a portion thereof accommodated in the body and being movable or rotatable to open or close a part or all of the flow paths or to change directions of the flow paths so that a flow of the fluid is adjusted.

Power conversion systems for converting thermal energy from a heat source to electricity are disclosed. In one exemplary embodiment, the power conversion system may include a substantially sealed chamber having an inner shroud having an inlet and an outlet and defining an internal passageway between the inlet and the outlet through which a working fluid passes. The sealed chamber may also include an outer shroud substantially surrounding the inner shroud, such that the working fluid exiting the outlet of the inner shroud returns to the inlet of the inner shroud in a closed-loop via a return passageway formed between an external surface of the inner shroud and an internal surface of the outer shroud. The power conversion system may further include a source heat exchanger disposed in the internal passageway of the inner shroud, the source heat exchanger being configured to at least partially receive a heat transmitting element.

The steam generating unit of dual circuit reactor with blowdown and drain system is implemented in the close loop, without any conventional blowdown expansion tanks and is designed for maximum pressure of the steam generator (SG) working medium. The SG blowdown water is combined into a single line, cooled down in the regenerative heat exchanger, then in the blowdown aftercooler and drain cooling line and taken out of the tight shell. Out of the tight shell, the SG blowdown water is supplied for treatment to the SG blowdown water treatment system designed for maximum pressure of the steam generator (SG) working medium. After treatment, the water returns to the tight shell and, via the regenerative heat exchanger, to the feed pipelines of each SG. The invention provides increased SG blowdown that leads to the accelerated chemical condition normalization even with considerable deviations.

Steam generators in power plants exchange energy from a primary medium to a secondary medium for energy extraction. Steam generators include one or more primary conduits and one or more secondary conduits. The conduits do not intermix the mediums and may thus discriminate among different fluid sources and destinations. One conduit may boil feedwater while another reheats steam for use in lower and higher-pressure turbines, respectively. Valves and other selectors divert steam and/or water into the steam generator or to other turbines or the environment for load balancing and other operational characteristics. Conduits circulate around an interior perimeter of the steam generator immersed in the primary medium and may have different cross-sections, radii, and internal structures depending on contained. A water conduit may have less flow area and a tighter coil radius. A steam conduit may include a swirler and rivulet stopper to intermix water in any steam flow.

Steam generators in power plants exchange energy from a primary medium to a secondary medium for energy extraction. Steam generators include one or more primary conduits and one or more secondary conduits. The conduits do not intermix the mediums and may thus discriminate among different fluid sources and destinations. One conduit may boil feedwater while another reheats steam for use in lower and higher-pressure turbines, respectively. Valves and other selectors divert steam and/or water into the steam generator or to other turbines or the environment for load balancing and other operational characteristics. Conduits circulate around an interior perimeter of the steam generator immersed in the primary medium and may have different cross-sections, radii, and internal structures depending on contained. A water conduit may have less flow area and a tighter coil radius. A steam conduit may include a swirler and rivulet stopper to intermix water in any steam flow.

Steam generator systems including tubesheet assemblies, such as for use in nuclear reactor systems, and associated devices and methods are described herein. A representative steam generator system can be installed in a nuclear reactor vessel positioned to house a primary coolant. The steam generator system can include a tubesheet assembly defining a plenum and comprising a tubesheet and a flexible connection portion coupling the tubesheet to the reactor vessel. The tubesheet can include a plurality of perforations fluidly coupled to the plenum. The steam generator system can further comprise a plurality of heat transfer tubes fluidly coupled to the perforations and configured to receive a flow of a secondary coolant. The connection portion can be more flexible than the tubesheet and the reactor vessel to reduce stresses on the tubesheet and the connections (e.g., tube-to-tubesheet (TTS) welds) between the tubes and the tubesheet during operation of the nuclear reactor system.