Abstract: A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.

Abstract: 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 fluid through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the 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 externally integrated once-through steam generator type small modular reactor, a steam generator is arranged along the circumference of a reactor vessel, and secondary cooling water flows in heat transfer tubes and changes to superheated steam. The small modular reactor includes: a nuclear reactor including a hemispherical upper head, the reactor vessel cylindrical shell coupled to the upper head and extending downward from the upper head in a cylindrical shape, and a hemispherical lower head provided on a lower portion of the reactor vessel cylindrical shell, wherein a core is placed in the nuclear reactor; the steam generator surrounding all around the reactor vessel cylindrical shell, the steam generator including a first penetration hole communicating with an inside of the nuclear reactor and a second penetration hole separate from the first penetration hole and communicating with the inside of the nuclear reactor.

Abstract: In an externally integrated steam generator type small modular reactor, a steam generator is arranged along the circumference of a reactor vessel cylindrical shell, and a steam drum is arranged along the circumference of the steam generator. The small modular reactor includes: a nuclear reactor including a hemispherical upper head, the reactor vessel cylindrical shell coupled to the upper head and extending downward from the upper head in a cylindrical shape, and a hemispherical lower head provided on a lower portion of the reactor vessel cylindrical shell, wherein a core is placed in the nuclear reactor; the steam generator surrounding all around the reactor vessel cylindrical shell and including a first penetration hole communicating with an inside of the nuclear reactor; and the steam drum surrounding the circumference of the steam generator and including a second penetration hole communicating with an inside of the steam generator.

Abstract: 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.

Abstract: A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.

Abstract: An integral pressurized water reactor (PWR) comprises: a cylindrical pressure vessel including an upper vessel section and a lower vessel section joined by a mid-flange; a cylindrical central riser disposed concentrically inside the cylindrical pressure vessel and including an upper riser section disposed in the upper vessel section and a lower riser section disposed in the lower vessel section; steam generators disposed inside the cylindrical pressure vessel in the upper vessel section; a reactor core comprising fissile material disposed inside the cylindrical pressure vessel in the lower vessel section; and control rod drive mechanism (CRDM) units disposed inside the cylindrical pressure vessel above the reactor core and in the lower vessel section. There is no vertical overlap between the steam generators and the CRDM units.

Abstract: An underwater electricity production module includes an elongated cylindrical casing, which includes an integrated electricity generation unit having a nuclear boiler. The generator is connected to an external electricity distribution station by electrical cables. The nuclear boiler is placed in a dry chamber of a reactor compartment associated with a chamber forming a safety water storage reservoir of the reactor. At least a radial wall of the reservoir chamber is in a direct heat exchange relationship with a marine environment that surrounds the underwater electricity production module in which the underwater electricity production module is submerged.

Abstract: A riser cone has a lower end sized to engage a cylindrical lower riser section of a nuclear reactor and an upper end sized to engage a cylindrical upper riser section of the nuclear reactor. The riser cone defines a compression sealing ring that is compressed between the lower riser section and the upper riser section in the assembled nuclear reactor. In some embodiments the riser cone comprises: a lower element defining the lower end of the riser cone; an upper element defining the upper end of the riser cone; and a compliance spring compressed between the lower element and the upper element. In some embodiments the riser cone comprises a frustoconical compression sealing ring accommodating a reduced diameter of the upper riser section as compared with the diameter of the lower riser section.

Abstract: This underwater electricity production module includes an elongated cylindrical box, which includes a reactor compartment and an electricity production unit. The reactor compartment includes a reservoir chamber and a dry chamber. A nuclear reactor is located in the dry chamber. The reservoir chamber forms a safety water storage reservoir. At least a radial wall of the reservoir chamber is in a direct heat exchange relationship with a marine environment that surrounds the cylindrical box. The nuclear reactor includes a nuclear boiler which includes a pressurizer connected through a depressurizing valve to the reservoir chamber.

Abstract: An underwater electricity production module includes an elongated cylindrical box, which includes a reactor compartment, which further includes a reservoir chamber and a dry chamber; an electricity production unit including a reactor container, the reactor container being placed in the dry chamber in a reactor pit having a first lower portion; and a circumferential wall including a first part delimiting the reservoir chamber, the first part being in a direct heat exchange relationship with water of the marine environment surrounding the elongated cylindrical box. The reservoir chamber has a second lower portion connected to the first lower portion through a water inlet duct placed along the circumferential wall. The reactor pit has a first upper portion connected to a corresponding portion of the reservoir chamber through a water return duct.

Abstract: The underwater electricity production module according to the invention includes means in the form of an elongated cylindrical box (12) in which means are integrated forming an electricity production unit including means forming a nuclear boiler (30), associated with electricity production means (37) connected to an external electricity distribution station by electrical cables, is characterized in that the nuclear boiler-forming means (30) are placed in a dry chamber (19) of the reactor compartment (18) associated with the chamber forming a safety water storage reservoir (20) of the reactor whereof at least the radial wall (53) is in a heat exchange relationship with the marine environment and in that the dry compartment (19) of the reactor container (18) is connected to the safety water storage reservoir chamber (20) of the reactor by depressurizing means (70) including means (71) forming a depressurizing valve placed in the upper portion of the dry chamber (19) and connected to one of the bubbler-forming m

Abstract: A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel having a lower portion containing a nuclear reactor core and a vessel head defining an internal pressurizer. A reactor coolant pump (RCP) mounted on the vessel head includes an impeller inside the pressure vessel, a pump motor outside the pressure vessel, and a vertical drive shaft connecting the motor and impeller. The drive shaft does not pass through the internal pressurizer. A central riser may be disposed concentrically inside the pressure vessel, and the RCP impels primary coolant downward into a downcomer annulus between the central ser and the pressure vessel. A steam generator may be disposed in the downcomer annulus and spaced apart from with the impeller by an outlet plenum, A manway may access the outlet plenum so tube plugging can be performed on the steam generator via access through the manway without removing the RCP.

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: Disclosed is an advanced process that relates to the enhanced production of energy using the integration of multiple thermal cycles (Brayton and Rankine) that employ multiple fuels, multiple working fluids, turbines and equipment. The method includes providing a nuclear reactor, reactor working fluid, heat exchangers, compressors, and multiple turbines to drive compressors that pressurize a humidified working fluid that is combusted with fuel fired in at least one gas turbine. The turbine(s) provide for electrical energy, processes or other mechanical loads.

Abstract: A nuclear reactor in which a primary coolant is contained, the primary coolant moves upwardly from the core by an operation thereof. An annular steam generator is arranged in an upper side of the core into which the upwardly moving primary coolant flows and transfers heat in the primary coolant into water therein to generate a steam. A passage structure defines a coolant passage for the primary coolant to an outside of the core. The heat-transferred primary coolant in the annular steam generator flows downwardly in the coolant passage so as to flow into the core, thereby moving upwardly. A reactor vessel is arranged to surround the coolant passage so as to contain the core, the annular steam generator and the passage means therein.

Abstract: A nuclear reactor cooled with pressurized water, having a pressurized tank installed in which are compact steam generators; each steam generator comprises a plurality of heat-exchange tubes having respective spiral portions set in levels on top of one another to form at least one annular tube bundle delimiting a substantially cylindrical internal central zone, pre-arranged for supply from above with primary water, which then traverses the tube bundle radially.

Abstract: The invention relates to an assembly for exchanging heat between first and second fluids, the assembly comprising a central manifold communicating with one of the inlet and the outlet for the first fluid; an annular manifold disposed around the central manifold and communicating with the other one of the inlet and the outlet for the first fluid; a plurality of heat exchangers interposed radially interposed between the central manifold and the annular manifold; and a plurality of axial inlet manifolds communicating with the inlet for the second fluid, and a plurality of axial outlet manifolds communicating with the outlet for the second fluid, the axial inlet and outlet manifolds being interposed circumferentially between the heat exchangers. According to the invention, the assembly has an inlet chamber disposed at a first axial end of the heat exchangers and putting the inlet(s) for the second fluid into communication with at least a plurality of axial inlet manifolds.

Abstract: A compact pressurised water nuclear reactor comprises a primary circuit fully integrated into the reactor vessel (10). Thus, a single steam generator (12) forms the cover of the vessel (10) and the pressuriser (30) and the primary pumps (28) are housed in the vessel (10). The same is true for the control mechanisms of the control rods (40). Finally, a venturi system (44) is also provided in the vessel (10) to create water circulation if there is a failure of the primary pumps (28).

Abstract: The invention relates to a pump (1) having a pump guide tube (13) for the transport of a heat-exchange medium (6) for a reactor having a bundle of catalyst tubes (2), where the pump (1) has a casing (14) which surrounds the pump guide tube (13), having an aperture (15) in the lower part of the casing (14) via which the heat-exchange medium (6) discharged from the lower region of the reactor by means of the pump (1) flows into the casing (14), flows upward in the region between the inner wall of the casing (14) and the outer wall of the pump guide tube (13), if desired via a heat exchanger (18), flows into the interior of the pump guide tube (13) via an aperture (16) in the upper region of the pump guide tube (13), flows through the pump guide tube (13) from top to bottom and flows via an aperture (17) in the lower region of the pump guide tube (13) into the reactor, into the upper region of the interspace between the catalyst tubes (2).

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving apertures formed therein. Each aperture has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: A nuclear reactor in which a secondary or tertiary coolant system of a nuclear steam supply system is simplified, comprising: a reactor vessel (2) which integrates a reactor core (1); a first coolant (3) which is stored in the reactor vessel (2) and heated by the reactor core (1) to convect; a first heat transfer tube (4) which is arranged in the reactor vessel (2) and comes into contact with the first coolant (3); and a second coolant (5) which is supplied from the outside of the reactor vessel (2) to the first heat transfer tube (4), cools the first coolant (3) and led to the outside of the reactor vessel (2).

Abstract: Disclosed is a forced cooling circular deep pool nuclear heating reactor with natural circulation comprising a water pool, a reactor core, a control rod, and a heat exchanger. The reactor core is placed on the bottom of the pool, the control rod is placed inside the reactor core, the heat exchanger outlet is connected with the pool water through the circulating pump, and its inlet is connected with the reactor core. Both the heat exchanger and the circulating pump are located outside the pool, and the depth of pool water is 10-60 meters. The circulating pump is an impeller-circulating pump with high specific speed. The nuclear reactor according to the present invention adopts forced circulation with natural circulation. Thus the passive conversion from forced circulation into natural circulation is realized and the parameters such as temperature of the reactor water will not rise abruptly and will change gradually.

Abstract: A fluid pump utilizing a canned rotor and canned stator is provided. The fluid pump has increased insulative properties over past “spool-type” pumps and has an increased ability to cool the stator, making it suitable for high temperature applications. A nuclear reactor is also provided. The reactor comprises a reactor vessel, that contains a nuclear fuel, control rods, reactor coolant and a reactor coolant pump for providing the reactor coolant to a steam generator. In a preferred embodiment, a steam generator is also provided inside the reactor vessel.

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving aperture formed therein. Each apertures has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: Passive emergency cooling in response to a loss of coolant accident (LOCA) in a PWR, having an integral reactor pressure vessel incorporating the steam generators and housed in a small high pressure containment vessel, is provided by circulating cooling water through the steam generators and heat exchangers in an external tank to cool the reactor vessel at a rate sufficient to lower the pressure in the reactor vessel below that in containment to reverse mass flow out of the reactor vessel and keep the reactor core covered without the addition of makeup water. Suppression tanks inside the small high pressure containment structure limit peak blowdown pressure in containment and provide flood-up water and gravity fed makeup water to cool the core. Diverse cooling is provided by natural circulation of air, and if needed, water, over the spherical containment structure.

Abstract: A fluid pump utilizing a canned rotor and canned stator is provided. The fluid pump has increased insulative properties over past “spool-type” pumps and has an increased ability to cool the stator, making it suitable for high temperature applications. A nuclear reactor is also provided. The reactor comprises a reactor vessel, that contains a nuclear fuel, control rods, reactor coolant and a reactor coolant pump for providing the reactor coolant to a steam generator. In a preferred embodiment, a steam generator is also provided inside the reactor vessel.

Abstract: In an indirect cycle nuclear reactor, a size of the reactor containment vessel is decreased by removing decay heat inside the reactor pressure vessel without using any active component to improve the economic feasibility. A main steam pipe communicating with a heat exchanger of the indirect cycle nuclear reactor is branched in a position upstream of a main steam isolation valve to connect the branched pipe to a heat exchanger in a pressure suppression pool through an isolation valve. A feed water pipe is also branched in a position upstream of an isolation valve to connect the branched pipe to the heat exchanger through the isolation valve. Decay heat is dissipated from the heat exchanger into the pressure suppression pool, and condensed water condensed by heat dissipation is returned to the heat exchanger to cool the inside of the pressure vessel.

Abstract: The present invention is used to reduce thermal load itself, being the cause to generate stress, which develops near liquid surface in a nuclear reactor wall and to contribute to further improvement of safety. A partition member (5) is arranged above a coolant liquid surface (9) in an annulus space (3) between a reactor vessel (1) and a guard vessel (2), a low-temperature gas is circulated through the annulus space above the partition member to cool down, the gas is circulated through the annulus space from under the coolant liquid surface to the partition member, and the high-temperature gas heated under the coolant liquid surface is used to raise the temperature above the coolant liquid surface.

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving apertures formed therein. Each apertures has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: A unitary, transportable, assembled nuclear steam supply system (NSSS) with a lifetime fuel supply utilizes a fast or epithermal spectrum reactor core immersed in a pool of light water together with a plurality of steam generators through which the coolant is circulated by up to 100% natural circulation at full power, augmented by reactor coolant pumps also immersed in the pool. Redundant steam generators and reactor coolant pumps, together with the fast or epithermal spectrum reactor core and pool configuration, make it possible to operate the NSSS for 10 to 15 or more years without maintenance on the internals or refueling, thereby rendering the system proliferation resistant.

Abstract: An apparatus for removing and installing a reactor component, such as a circulating pump, by remote operation. The apparatus includes a handling mechanism for transporting various interchangeable containers and tools along a path between a position on top of the suppression pool and a position overlying the reactor component. The handling mechanism has a yoke assembly adapted to receive and support the trunnions of any container or tool, lifting screws for raising and lowering the yoke assembly and an arcuate rack mounted on the shield wall for rotating the yoke assembly around the circumference of the shield wall. A transition collar is latched onto the pump pressure housing. The transition collar has a valve at the top for closing off its interior. Seals arranged between the transition collar and pressure housing to ensure water-tightness.

Abstract: A fast reactor comprises a core composed of nuclear fuel, a core barrel surrounding an outer periphery of the core, an annular reflector surrounding an outer periphery of the core barrel, a partition wall structure surrounding an outer periphery of the annular reflector and supporting the core barrel by a supporting structure arranged radially of the fast reactor, the partition wall structure constituting an inner wall of a coolant passage for a primary coolant, a neutron shield surrounding an outer periphery of the partition wall structure and disposed in the coolant passage, a reactor vessel surrounding an outer periphery of the neutron shield and having an inner wall constituting an outer wall of the coolant passage, and a guard vessel surrounding an outer periphery of the reactor vessel.

Abstract: A pressured-water nuclear reactor comprising an apparatus for directly removing the residual power from the core of the reactor is provided. Advantageously, upon shut-down of the nuclear reactor of the present invention, residual power of the reactor may be safely removed therefrom.

Abstract: A pressurized-water nuclear reactor with inherent safety comprises a pressurized vessel filled with borated water within which is the body of the reactor. The latter includes an upper part and a lower part within which is the reactor core. The upper part includes a first tubular cavity, a second tubular cavity coaxial with and outside the first cavity and a plurality of helical tube bundles extending between respective pairs of connectors for the inlet and outlet of water circulating in the tube bundles.

Abstract: A tank type nuclear reactor comprises an outer shielding wall structure, a reactor vessel installed in the shielding wall structure and provided with an inner sealed space in which a primary liquid coolant circulates, a reactor core disposed in the reactor vessel and charged with a number of fuel assemblies, a core support structure disposed to a bottom of the reactor vessel, a partition wall structure disposed on the core support structure defining the inner space of the reactor vessel into a hot pool region and a cold pool region, intermediate heat exchangers disposed in the reactor vessel for carrying out heat exchanging operation between the liquid coolant flow from the hot pool region and a secondary side heat medium, and electromagnetic pumps disposed in the reactor vessel for feeding the primary coolant into the core. The partition wall structure has a vertically cylindrical wall structure including a lower portion constituting a core tank portion surrounding an outer periphery of the core.

Abstract: A liquid metal cooled nuclear reactor for a nuclear power plant on the modular principle is disposed in a reactor cavern having cooling surfaces. A reactor tank contains a reactor core and one or more heat exchangers and primary pumps and is surrounded by a double tank of nodular graphite cast iron serving as a heat accumulator and having a plurality of detachably interconnected and superposed rings and a base. With this configuration, not only is the structure simplified but advantages are also achieved for operation, inspection and repair or exchange. Producing the double tank from cast iron has appreciable advantages for the construction of the entire plant with respect to external influences. The function as a heat accumulator has appreciable advantages for the construction of the heat removal systems of the cooling surfaces.

Abstract: An expanded reactor core capacity is achieved within the constraints imposed by the diametric extent of the current advanced boiling reactor (ABWR). This enhanced core capacity is achieved through the utilization of a scalloped shroud structure in conjunction with modified control rod designs. The core diameter is expanded toward the reactor vessel interior wall to establish a minimum distance therebetween representing a minimum value thereof avoiding neutron fluence induced vessel embrittlement. The control rods servicing peripheral fuel assemblies are configured having either a T-shaped blade configuration or an L-shaped blade configuration depending upon the parallel orientations of the peripheral fuel assemblies.

Abstract: A recirculation system is disclosed for driving reactor coolant water contained in an annular downcomer defined between a reactor vessel and a reactor core spaced radially inwardly therefrom. The system includes a plurality of circumferentially spaced pumps disposed in the downcomer, each pump including an inlet for receiving coolant water from the downcomer as pump inlet flow, and an outlet for discharging the pressurized water. The recirculation system firstly increases the pressure of the pump inlet flow at the pump inlet before being sucked into the pump for being further pressurized by the pump.

Abstract: An intrinsic-safety nuclear reactor of the pressurized water type, having:a reactor vessel (2) equipped with a core (4), a lower header (5) and an upper header (6), at least one heat exchanger (3) with a secondary fluid, means of hydraulic connection (9, 10) between said headers and said heat exchanger and at least one circulation pump (11),a pressurized container (1) surrounding the reactor vessel (2) and which defines a tank (15) full of a cold, neutron-absorbing liquid;pipes (20) allowing communication between the lower area of said tank and the lower header of the vessel, as well as pipes (21) allowing communication between the upper area of said tank and the upper header,in which the pressure drop in the primary fluid across the core is substantially equal to the difference in head between the cold column of said tank and the hot column of the vessel.The pressurized container (1) is immersed in a pool (18) containing a neutron-absorbing liquid at atmospheric pressure.

Abstract: A water cooled nuclear reactor comprises a reactor core, a primary water coolant circuit and a pressurizer arranged as an integral unit in a pressure vessel. A passive full pressure emergency core cooling and residual heat removal system is provided which comprises a tank having a reserve supply of water positioned above the primary water coolant circuit or the reactor core. The tank is interconnected to the primary water coolant circuit by a first pipe and a second pipe. The first pipe has an inverted U-bend which passes through a water space and a steam space of the pressurizer to form a vapor lock. The first pipe and second pipe are also provided with hydrostatic thermal seals. The tank has one or more residual heat removal circuits comprising a heat exchanger positioned in the tank and a heat exchanger outside the tank. Movement of the vapor lock from the inverted U-bend allows cool water from the tank to flow into the primary water coolant circuit and hot water to flow into the tank to be cooled.

Abstract: A water cooled nuclear reactor comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper chamber and a lower chamber by a casing, the reactor core and primary coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A movable diaphragm is positioned in the upper chamber, and is sealingly secured to the casing by a bellows arrangement to divide the upper chamber into a water filled space and a gas filled space. A plurality of surge ports interconnect the water space with the primary coolant circuit. The diaphragm moves to accommodate changes in the volume or pressure of the water in the primary coolant circuit and water space. The diaphragm is loaded by springs and dampers to prevent oscillation of the diaphragm. Alternatively the diaphragm may be an elastic membrane.

Abstract: A recirculation system for a boiling water reactor includes a plurality of circumferentially spaced impeller-driven reactor internal pumps disposed in a downcomer for pumping a first portion of reactor coolant, and a plurality of circumferentially spaced fluid-driven jet pumps disposed in the downcomer for pumping a remaining portion of the coolant in the downcomer. In an exemplary embodiment, the jet pumps are driven by a portion of feedwater provided to the reactor.

Abstract: A recirculation system is disclosed for driving reactor coolant water in an annular downcomer defined between a reactor vessel and a core shroud spaced radially inwardly therefrom. The system supplies feedwater to the vessel and to a turbopump disposed inside the downcomer. The turbopump in accordance with one embodiment of the present invention includes a stationary axle and a pump impeller rotatably joined thereto and having an inlet end for receiving the coolant water from the downcomer. An annular plenum surrounds the impeller for channeling feedwater to a plurality of turbine blades joined to the impeller for rotating the impeller for driving the coolant water. The impeller is lubricated solely by the feedwater upon rotation of the impeller about the axle.

Abstract: In order to reduce any loss of primary water coolant from around a reactor core of a water cooled nuclear reactor caused by any failure of a pressure vessel, an inner vessel is positioned within and spaced from the pressure vessel. The reactor core and main portion of the primary water coolant circuit and a heat exchanger are positioned within the inner vessel to maintain some primary water coolant around the reactorcore and to allow residual decay heat to be removed from the reactor core by the heat exchanger. In a second embodiment an aperture at the upper region of the inner vessel is dimensioned configured and arranged to prevent steam from a steam space of an integral pressurized water cooled nuclear reactor for a ship entering the main portion of the primary water coolant circuit in the inner vessel if the longitudinal axis of the nuclear reactor is displaced from its normal substantially vertical position to an abnormal position at an angle to the vertical direction.

Abstract: An autonomous, decentralized fast breeder reactor includes a single reactor main vessel which houses a plurality of small-size reactor subsystems each having a small-scale fast breeder reactor core, and a plurality steam generator subsystems. These subsystems function in an autonomous manner and are caused to undergo a heat transfer with one another by a coolant circulating naturally through the interior of the main vessel, thereby constructing a cooperatively operating system. Steam generated by the steam generators is introduced to a turbine system and utilized in generating electricity. The condensate from the turbines is cooled by a heat accumulating pool, and the heat is utilized in a separate system. The entire system is installed underground and use is made of the difference in elevation. Use also is made of solid bedrock to construct a housing facility for the reactor.

Abstract: A nuclear reactor includes a reactor vessel, a reactor core disposed within the vessel and including fuel rods, a primary cooling system disposed in the vessel and including a steam generator and a pump and a poison tank disposed within the vessel and containing high concentration boric acid solution. The poison tank surrounds the reactor core, and the steam generator surrounds a portion of the poison tank.

Abstract: In a high-temperature gas-cooled nuclear reactor system, which is suitable for modular utilization in combination with one or more other similar nuclear reactor systems, an arrangement which comprises passive heat sink means for absorbing decay heat energy generated at a reactor core included within said reactor system; and means for removing heat energy from said heat sink means at a rate sufficient to maintain the capacity of the heat sink means for absorption of decay heat energy such that release of fission products resulting from loss of forced circulation of gas coolant, or such loss in combination with coolant depressurization, when the reactor is critical and at power is prevented, thereby to provide safety means independent of human or automatic activation; and method embodiments corresponding to same.

Abstract: A nuclear power system comprises a plurality of modules disposed in below-grade pits to provide a compact, self-contained nuclear power supply. The modules are preferably individually transportable so that they may be substantially preassembled prior to installation. The system operates at relatively low temperatures and pressures, and includes various safety features which would prevent radioactive contamination of the surrounding environment in the event of a disturbance causing rupture of one or more of the odules or the pipes interconnecting the modules. The system also provides a low resistance flow path for vapor discharged from the turbine to improve efficiency.