Abstract: A reactor containment vessel comprises typically a reactor pressure vessel housed in a dry well of the reactor containment vessel, a vent passage through which steam in the dry well is introduced into coolant accommodated in a pressure suppression chamber of the reactor containment vessel, a closed space formed at a position lower than the level of the normal liquid surface of the coolant, a first passage having an inlet opened into the pressure suppression chamber at a level higher than that of the normal liquid surface of the coolant and an outlet opened into the closed space, and a second passage communicating between the closed space and the dry well through a counter flow preventing arrangement for checking the flow directed toward the closed space, and therefore, noncondensable gas and liquid accumulated in the pressure suppression chamber are discharged into the dry well the pressure in which is higher than that in the pressure suppression chamber by making use of a difference in water head increased d

Abstract: The invention is directed to a natural circulation reactor having a reactor pressure vessel with a core housed therein, the core being disposed in such a location that a top portion of the core is submerged under coolant even in the event that any pipe connected to the reactor pressure vessel is broken and then a coolant level in the reactor pressure vessel is lowered due to flushing. This permits the reactor core to be submerged under coolant even in the event of breakage of any pipe connected to the reactor pressure vessel, ensuring to eliminate a possibility that the top portion of the reactor core is exposed temporarily during an intermediate period before actuation of an accumulated coolant injection system to start injecting of the coolant into the reactor pressure vessel after the end of flushing.

Abstract: A nuclear reactor facility including a primary containment vessel, a reactor pressure vessel installed in the primary containment vessel and accommodating a reactor core in a lower part thereof, and a vertical cylindrical wall disposed in a lower part of the primary containment vessel around and spaced from the reactor pressure vessel so as to delimit an annular space therebetween. The vertical cylindrical wall has an upper end disposed at a position higher than an upper end of the reactor core and a diaphragm extends substantially horizontally between the upper end of the vertical cylindrical wall and an inner wall of the primary containment vessel for cooperating with the vertical cylindrical wall to separate a space in the primary containment vessel around the reactor pressure vessel into a pressure suppression chamber and a drywell which includes annular space.

Abstract: A nuclear system of the type which includes a containment wherein a nuclear reactor pressure vessel is located has a suppression pool to which steam can be vented so that it will condense and reduce the pressure in the pressure vessel upon the happening of an accident such as loss of reactor coolant or a steam pipe failure. Steam also can be vented directly to the containment space to further reduce pressure in the reactor. When reactor pressure is lowered to a certain pressure value, a gravity supply of water from an elevated pool of water will have a sufficient head to flow against the pressure in the reactor and into the reactor to submerge the fuel rods in the reactor. One or more isolation condensers are submerged in a large supply of water this supply being elevated some distance above the pressure vessel. At least one isolation condenser has inlet thereto communicated to an open entry conduit disposed in the containment.

Abstract: The invention refers to a solution for the reactor block of a sodium-cooled fast reactor, fitted with an inner cylindrical vessel. This forms the mechanical structure for hydraulic separation 7 between the primary sodium hot header 2 and the cold header 3. The hot header is located in the area above the core, and the cold header, having a larger volume and an annular shape, is located all around the hot header. According to the invention the primary pumps and intermediate exchangers 8 are immersed in the cold header 3. The intermediate heat exchangers 8 are of the straight tube bundle type, with primary sodium outside the tubes, and they are hydraulically connected to the hot header by a transverse duct 13, solid with the inner cylindrical vessel 7, and mechanically sealed where it interfaces with the heat exchanger. This mechanical sealing system is made as close against the heat exchanger, as possible, so that the heat exchanger is removable.

Abstract: A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

Abstract: High temperature reactor with residual-heat transfer system comprises a cooling gas intake at the bottom and cooling gas outlet at the top so that a cooling gas can flow from the bottom to the top through the reactor core. In order to assure reliable heat transfer a bypass duct is provided with a lower end communicating with the cooling gas intake and the upper end communicating with the cooling gas outlet. The bypass duct is arranged parallel to the reactor core and passing a partial flow of cooling gas from the bottom to the top. This partial flow of cooling gas heats up only trivially. This partial flow of cooling gas is further cooled by the cooler. The upward flow of the comparatively cold cooling gas in the bypass duct stops and by itself reverses because the cooling gas in the bypass duct is drawn toward the reactor core on account of the natural convection.

Abstract: A nuclear reactor system which includes a containment uses, upon loss-of-coolant event, an isolation condenser submerged in a large supply of water and elevated some distance above the system pressure vessel to effect both initial and decay heat dissipation cooling in the containment. The isolation condenser has inlet thereto communicated to an open entry conduit disposed in the containment so that steam and heated gasses in the containment space enter the isolation condenser and are cooled. Condensate resulting from the cooling is returned to an elevated system gravity coolant supply pool, which pool is used for replenishing coolant lost from the pressure vessel, the return being through a return conduit that has a lower end section configured with a water trap with non-condensable gasses present in the steam being separated from the condensate and vented to the suppression pool.

Abstract: A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

Abstract: A nuclear fuel element for a particle bed reactor. Concentric inner and outer porous cylinders attached at either end define an annulus therebetween that contains a fuel particle bed of nuclear fuel material. The outer porous cylinder is provided with fins that extend radially outward along its entire length. A nonporous cylinder is attached to the outer porous cylinder at either end and is positioned around the outer porous cylinder so as to be concentric therewith. Flexible fins extend radially inward form the nonporous cylinder along its entire length and contact the fins on the outer porous cylinder for conduction of heat to heat removal tubes on the outer surface of the nonporous cylinder. The heat removal tubes allow bimodal cooling of the particle bed reactor.

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: A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Abstract: A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Abstract: A primary containment vessel, comprising a reactor pressure vessel in which a core is accommodated, a dry-well volume covering the upper portion of the reactor pressure vessel, a suppression chamber provided with a suppression pool surrounding an intermediate portion of the reactor pressure vessel, has a plurality of vent tubes establishing a connection between the dry-well volume and the suppression pool and an annular baffle plate disposed in the suppression pool along the side wall of the suppression chamber, having its top end positioned at a height between a free surface of the suppression pool and an outlet of the vent tube opened in the suppression pool and its lower end positioned at a height between the outlet and the bottom of the suppression pool. If a loss of coolant accident occurs, hot steam discharged into the dry-well volume is discharged into the suppression pool through the vent tube so as to heat water on the inside of the baffle plate.

Abstract: A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

Abstract: A container-outer-periphery pool in which water is stored is provided between a primary containment vessel and a reactor building, and outside of a pressure suppression pool. Consequently, the natural heat transfer capabilities of reactor facilities is enhanced, and the inherent safety of the reactor facilities is improved.

Abstract: A heat-generating member (3), in particular a nuclear reactor core, is placed in a liquid contained in a pressure vessel (1) and adapted to serve as a coolant for the heat-generating member. For cooling of the liquid the pressure vessel is adapted to be included in a circulation system (15a, 15, 14. 16, 16a) for self-circulation of the liquid and/or of steam of the liquid with any contents of uncondensable gas. The circulation system also includes an evaporator (14) arranged in an evaporation pool (13), a supply conduit (15) for conducting liquid and/or steam from a point of connection (15a) in the upper part of the pressure vessel to the evaporator, and a discharge conduit (16) for conducting liquid from the evaporator to a point of connection (16a) on the pressure vessel which is located below the point of connection (15a) for the supply conduit. The evaporator is located at a higher level than the point of connection (16a) on the pressure vessel for the discharge conduit (16).

Abstract: An improved emergency cooling system is provided for an organic cooled and moderated nuclear reactor. The cooling system permits an inherently safe reactor design to be achieved having a number of other novel and significant advantages. Most importantly, the reactor can be designed to survive a largest credible accident which involves the loss of the entire primary coolant inventory, while continuing to provide core cooling in a passive mode for a period of one or more days post-accident.

Abstract: A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Abstract: A novel liquid nuclear reactor is described which comprises a reactor vessel that is connected through upper and lower liquid conduit means to one or more satellite tanks that contain a heat exchanger means and pump means.

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.

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: Gases and entrained particulate matter are automatically vented from a nuclear containment as pressure approaches a set point of the containment design during a postulated accident scenario. A porous bed filter is submerged in a tank of water, affording a first stage of filtration of the vented gases and entrained particulate matter passing in an upward flow from the bottom of the tank through the porous filter bed and a second, pool scrubbing stage as the flow proceeds upwardly through a central region of the water above the filter with accompanying decay heat removal prior to release of the gases from the water surface and discharge to atmosphere. The water returns through a down-flow path defined by an annular boundary region surrounding the central region of the water volume, in a pumping function for repeating the mixing and two-stage flow, and purges the porous filter bed.

Abstract: The emergency core cooling structure according to the present invention is provided with a water storage container which holds therein emergency cooling water to be supplied to a core region in a nuclear reactor in case of an emergency, for example, when the primary cooling water flows out from a reactor vessel, so as to maintain the temperature of the emergency cooling water in the upper portion of the water storage container in a level higher than the level of that of the emergency cooling water in the lower portion thereof, the lower portion of the water storage container, which is filled with the low-temperature cooling water, being communicated at all times with the core of the nuclear reactor by a means which has substantially no movable member.

Abstract: A nuclear power plant with a gas cooled high temperature reactor, installed in a prestressed concrete pressure vessel with the operational and decay heat removal systems. The prestressed concrete pressure vessel has a thermal protection system, with a thermal insulating layer and a liner cooling system. The liner cooling system, which is includes water carrying cooling pipes, which along with intermediate heat exchangers and cooling water pumps make up a closed intermediate cooling loop used for removal of the decay heat in case a failure of the decay heat removal systems. The elements of the invention assure an adequate water flow for the removal of decay heat in the liner cooling system in any situation, i.e. such that the decay heat may also be removed by natural convection. These element insure a sufficient driving pressure differences and minimize pressure losses in the intermediate cooling loop.

Abstract: A nuclear power plant with a high temperature reactor located eccentrically in a prestressed concrete pressure vessel with spherical fuel elements for a capacity of 100-300 MWth intended primarily for power generation. The plant is inherently safe even in extreme accidents due to special devices for removal of decay heat. It is provided that upon a failure of the operational decay heat removal devices, decay heat is removed through a liner cooling system of the prestressed concrete pressure vessel. The liner cooling system is connected to a water-filled elevated reservoir by an external circulation loop. The elevated reservoir is connected to a further heat sink through a recooling circulation loop. It is possible to feed water into the elevated reservoir.

Abstract: The cooling device comprises a heat exchanger (7) immersed in the hot header (4) of the nuclear reactor vessel, a substantially vertical tubular conduit (14) passing through the stepped wall (2) vertically below the exchanger (7), a bell (16) fixed under the exchanger (7) in the extension of its lower part and a mechanism for compressing inert gas and relieving pressure in the internal space of the bell (16). The inert gas compressed in the bell (16) enables the liquid metal in the hot header (4) to be completely separated from the liquid metal in the cold header (5). When the pressure in the bell (16) is relieved, the outlets (10) for the cooled liquid metal in the exchanger (7) are brought into communication with the cold header (5) through the bell (16) and the conduit (14).

Abstract: An improved decay heat removal system and apparatus for a nuclear boiling water reactor. The apparatus includes an isolation condenser shell, water coolant, and heat exchange surfaces that are immersed in the water coolant and are sized for shutdown cooling duty. The apparatus is time shared for both isolation cooling and shutdown cooling duty. The invention reduces the total number of heat exchange surfaces and heat exchanger shells required for dissipating heat generated by the reactor core as compared to conventional, separate isolation cooling and shutdown cooling systems. In addition, the amount of reactor building space which must be reserved for isolation cooling and shutdown cooling requirements may be significantly reduced.

Abstract: A low capacity nuclear reactor with spherical fuel elements laid out in an underground configuration and characterized by a compact structure and the far-reaching elimination of active operating installations, such as a charging apparatus, gas purification installations and control systems. The reactor is particularly suitable for generation of heat for heating purposes. The graphite reflector surrounding the pile of fuel elements on all sides includes layers of spherical graphite elements with a diameter equal to that of the fuel elements. The poured part of the graphite reflector and the pile of fuel elements are located in a metal core vessel made of lattice work or perforated sheet metal and capable of supporting the entire weight of the graphite and fuel elements. The mesh or the holes of cage-like core vessel are smaller than the diameter of the spherical elements.

Abstract: A shroud tank and fill pipe for a boiling water nuclear reactor of the type having a reactor vessel containing a saturated water coolant inventory, a reactor core for heating water to generate a steam/water mixture, a steam separator for classifying said steam and water from the steam/water mixture, and standpipes for conveying the steam/water mixture from the core to the steam separator. The shroud tank is disposed inside the reactor vessel overlying the reactor core and circumscribing the standpipes. The shroud tank has an open top, a closed bottom, and a plurality of drain holes around its bottom periphery. During normal reactor operation, the shroud tank is supplied with cold water coolant which is continuously introduced at relatively low flow rates into the shroud tank through the fill pipe. The drain holes are sized to minimize the outflow of cold water coolant during normal operation.

Abstract: Two systems of low-pressure core spray appratuses have coolers, and are adapted to supply cooling water to a core spray header provided in a core-surrounding cylindrical shroud in a reactor pressure vessel. Two systems of high-pressure core flooding apparatuses and one system of high-pressure coolant injection apparatus are adapted to supply the cooling water to a region formed between the shroud and reactor pressure vessel. The elevation of the openings, which are in the reactor pressure vessel, of the high-pressure core flooding apparatuses and high-pressure coolant injection apparatus are higher than that of the core spray header. A pipe for returning the cooling water in the reactor pressure vessel to the above-mentioned coolers is connected to either the portion of the interior of the reactor pressure vesssel which is below the core or the portion of the interior of the reactor pressure vessel which is between the walls of the shroud and reactor pressure vessel.

Abstract: Device for condensing steam under pressure consisting of a unit incorporating a storage vessel (6) containing cooling water, a distribution and heat exchange unit (15) fixed inside the storage vessel (6) and a water supply (30,31) for replacing the water in a storage tank (6) which is vaporized by contact with the tubes (24) of the distribution and heat exchange unit (15). A stack (7) is connected to the upper part of the storage vessel (6) and a tranquilizer grid is arranged in the storage vessel, above the heat exchange unit (15). The circulation of the fluid which is constituted of a two-phase mixture of water and steam is thus activated, while avoiding drawing out a significant quantity of water with the steam.

Abstract: Gas-cooled high-temperature nuclear reactor having a reactor core comprising individual fuel elements provided with means for forming a barrier against the release of fission products producible therein during reactor operation, the fuel elements being received in a cylindrical barrel formed of an inner graphite layer functioning as a reflector, an outer layer of insulating material surrounding the inner layer, and a metallic receptacle, the inner and outer layers and the receptacle being formed of respective side, bottom and cover portions, the side and cover portions of the inner layer being formed with first channels into which means for controlling the reactors are insertable, the bottom, side and cover portions of the inner layer being further formed with second channels wherein, during reactor operation, cooling gas is circulated under pressure from the bottom to the top of the receptacle, the bottom portion of the inner layer having first openings for introducing cooling gas into the second channels du

Abstract: A heat radiator useful for expelling waste heat from a power generating system aboard a space vehicle is disclosed. Liquid to be cooled is passed to the interior of a rotating bubble membrane radiator, where it is sprayed into the interior of the bubble. Liquid impacting upon the interior surface of the bubble is cooled and the heat radiated from the outer surface of the membrane. Cooled liquid is collected by the action of centrifical force about the equator of the rotating membrane and returned to the power system. Details regarding a complete space power system employing the radiator are given.

Abstract: A concrete reactor pressure vessel for a low capacity, gas cooled nuclear reactor is provided. To assure the safe containment of the cooling gas in an economical manner and to remove the heat generated in the reactor core, the pressure vessel is equipped with reinforcing or prestressing elements. Cooling gas transfers its heat to a heat exchanger means consisting of the liner of the pressure vessel and cooling pipes mounted on the concrete side of the liner with water flowing through the cooling pipes. The heat exchanger comprises on the secondary side at least two mutually independent systems, each provided with collector pipe means located outside the reactor pressrue vessel for the forerunnings and afterrunnings of the cooling water. Each cooling pipe is connected by means of an inlet and a drain line with the corressponding collector pipe means.

Abstract: A liquid metal cooled pool type nuclear reactor incorporates an emergency heat exchanger having a novel geometry. The heat exchanger comprises circular and annular tube plates coaxially aligned. The tube bundle has a vertical straight part connected to the central tube plate, a bent horizontal circular portion extending over one-third of the heat exchanger's circumference for returning the bundle, and a vertical straight return part joining the peripheral tube plate.

Abstract: A pressurized water nuclear reactor has a reactor vessel arranged in a pool, which is filled with a neutron absorbing liquid, for example borated water. The reactor vessel is closed except for tubes connecting it with a tray above it. The coolant in the circuit rises from the vessel to the tray, gives up its heat by flashing, and flows back to the bottom of the vessel, driven by natural circulation. The tray is separated from the pool by a vapor-filled bell, which surrounds it. In the bell the vapor gives up its useful heat to a condenser. The relatively low boron content of the cooling circuit, compared to the pool, is achieved by continuous dilution of the condensate from vapor additionally generated out of the pool water. The dilution process is an equilibrium with continuous inflow of the pool water. The inflow is automatically controlled by the pool level, which rises when the pool water is pressed out from below the bell by overproduction of vapor.

Abstract: An improved shut-down heat removal system for a liquid metal nuclear reactor of the type having a vessel for holding hot and cold pools of liquid sodium is disclosed herein. Generally, the improved system comprises a redan or barrier within the reactor vessel which allows an auxiliary heat exchanger to become immersed in liquid sodium from the hot pool whenever the reactor pump fails to generate a metal-circulating pressure differential between the hot and cold pools of sodium. This redan also defines an alternative circulation path between the hot and cold pools of sodium in order to equilibrate the distribution of the decay heat from the reactor core. The invention may take the form of a redan or barrier that circumscribes the inner wall of the reactor vessel, thereby defining an annular space therebetween.

Abstract: An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged.

Abstract: Intrinsically safe emergency cooling device for a pressurized-water nuclear reactor, comprising an auxiliary circuit which feeds the steam generator and in which is arranged a condenser (10) receiving the steam from the generator and condensing it. The condensate is returned to the steam generator by means of gravity. The device also incorporates a passive adjustment means comprising an adjustable valve (21), a cylinder (23), the chamber of which communicates with the steam pipe (7) and contains a piston (33), and a coupling means (32) between the piston (33) and the shut-off element (30) of the valve (21). The feed rate of the steam generator and the cooling power of the condenser (10) can be regulated in this way.

Abstract: A nuclear power plant (heating plant) with a helium cooled high temperature reactor with spherical fuel elements, located in a cylindrical prestressed concrete pressure vessel, suitable for supplying heat (for local or remote heating or the generation of process steam) with a capacity of approximately 50 to 300 MWth. The high temperature reactor core is located out of center in the prestressed concrete pressure vessel, with at least two heat exchangers installed adjacent and offset in the upward direction relative to it. The heat exchangers are each connected on the secondary side to an intermediate circulation loop including an intermediate heat exchanger and circulating pump, and have an auxiliary circulation loop with a recooling system connected in parallel with them for the removal of decay heat. The auxiliary loops are closed in normal operation.

Abstract: A nuclear reactor installation located in the cavity of a pressure vessel comprises a nuclear reactor with a core traversed from top to bottom by a cooling gas, a plurality of main loops consisting of heat exchangers and blowers, together with auxiliary loops for the removal of decay heat. According to the invention, the auxiliary loops contain no heat exchangers but a bundle of heat pipes independent of each other, the heat absorbing part of which is arranged in the pressure vessel cavity. As a heat sink, an external cooling water loop is provided for each bundle of heat pipes in which water is circulated. The cooling water is recooled in a cooling tower. In a preferred embodiment the heat transferring part of the heat pipes of each bundle terminates in an external water reservoir to which the cooling water loop is connected. For a certain period of time, in case of a sufficient volume of water, the decay heat may be removed only by evaporation from the water reservoir.

Abstract: A nuclear reactor includes a core submerged in a pool of liquid. Under normal conditions, coolant flows through the core without intermixing with the liquid in the pool. In the event of failure of the primary coolant circulation system, liquid from the pool flows through openings in the primary circulation system so as to cool the core by natural convection. Flow through the openings during normal operating conditions may be controlled regardless of the flow rate.

Abstract: A single loop nuclear power plant with a helium cooled high temperature reactor for generation of electric current, designed for a capacity of 1-5 MWe. The plant, which in addition to the high temperature reactor includes a gas turbine assembly and a heat exchange apparatus, is housed in two pressure vessels located above each other and connected in a releasable manner. The lower pressure vessel contains the high temperature reactor and is charged with the primary gas. The other circulation components are located in the upper pressure vessel which is filled with a protective gas. The gas turbine, the radiators, the high temperature compressor, the intermediate radiators, and the low pressure compressor, are arranged above each other in this sequence and aligned with the high temperature reactor. the recuperator is laterally arranged. A generator may also be located in the upper pressure vessel or in a container set upon the upper pressure vessel.

Abstract: The invention relates to a nuclear power station for a gas cooled high temperature pebble bed nuclear reactor. The nuclear power station is characterized by a combination of features, whereby the system inherent properties of a high temperature reactor are utilized to make possible the economical operation of a nuclear power station of medium capacity (300-600 MW.sub.el) while maintaining a high standard of safety. The characteristics comprise a reactor protection building equipped with pressure relief means in combination with filters, several auxiliary cooling systems separate from the operating cooling systems for the removal of decay heat in the case of accidents, and the utilization of a liner cooling system for the prestressed concrete reactor pressure vessel to assure the removal of the decay heat in case of a failure of the auxiliary cooling systems.

Abstract: A passive safety system for a nuclear reactor is comprised of a first subsystem for circulating water solely by natural convection from a first branch to a second branch of a reactor coolant circuit with the circumvention of a steam generator for removing decay heat from the reactor coolant circuit at any pressure. The first branch guides heated water from the reactor vessel into the steam generator and the second branch guides cooled water from the steam generator into the reactor vessel. The first subsystem includes a heat exchanger for cooling water flowing from the first branch and prior to being introduced into the second branch and a first valve for allowing flow of water from the first branch into the second branch solely in response to a parameter value pertaining to operational safety. There is also provided a second subsystem for introducing stored cold water solely by gravity into the reactor vessel for making up for lost water in the reactor coolant circuit at any pressure.

Abstract: A double tube modular coil steam generator is provided in which a multiplicity of inner tubes conducting water are individually surrounded by outer tubes containing liquid metal as a heat transfer agent. The double tubes form into coils, providing a large surface area while conserving space. Immersion of the double tube coil in hot liquid metal, e.g., from the core of a nuclear reactor, causes efficient transfer of heat across the liquid metal in the outer tube to the water in the inner tube, creating superheated steam, which can be cycled to a turbine and converted to electrical power.The efficiency, reliability and safety of the multiple double tube design of the steam generator obviates the necessity of many secondary heat removal and emergency components in addition to conserving space and material. The modular design allows ease of operation, fabrication and repair.

Abstract: A gas cooled high temperature nuclear reactor utilizes an independent cooling system for the safety enclosure surrounding the reactor vessel. The cooling system comprises means for circulating cooling medium at least on the reactor side of a concrete safety enclosure shell and a separate closed cooling loop for circulation of separate cooling medium through a reservoir of the first cooling medium and to the outside of the entire nuclear reactor installation.

Abstract: A nuclear reactor installation with a high temperture, gas cooled nuclear reactor supported in a cavity on a segmented bottom plate held by a centered point of fixation. The first heat exchangers of the primary circulation loop are arranged in the cavity around the nuclear reactor, with the associated first blowers being supported in first passages of the pressure vessel roof. The second heat exchanger for the removal of the decay heat are arranged similarly, with the associated second blowers located in passages of the pressure vessel bottom. The blowers are equipped with contactless magnetic bearings for the rotors. The drives of the core rods and the reflector rods are designed for a banking operation. Simultaneously, the side reflector is supported by means of radial supports on the thermal side shield elastically and secured against rotation. The measures described provide a simplified configuration without affecting safety.

Abstract: A nuclear reactor is provided characterized by a circulating liquid metal cooling system comprising a heat exchanger having a closed intermediate heat transfer fluid circuit in the form of a helical coil juxtaposed with the liquid metal coolant system, which effectively transfers heat through the intermediate circuit to a secondary fluid circuit. The intermediate heat transfer fluid circuit, which completely separates the liquid metal cooling system from the secondary fluid circuit, prevents potentially dangerous reactions between the primary liquid metal coolant and the secondary fluid (e.g., water).The efficiency, reliability and safety of the heat exchanger features eliminates the need for many secondary heat removal and emergency components in the design of the nuclear reactor.