Abstract: A reactor core of a fast breeder reactor is supported by a core support within a reactor vessel. The reactor vessel is provided with a first or lower sodium chamber for accommodating low temperature sodium, a second or intermediate sodium chamber for accommodating low temperature sodium and a third or upper sodium chamber for accommodating high temperature sodium heated by passage of the low temperature sodium through the reactor core into the third sodium chamber. Low temperature sodium is introduced into the second sodium chamber arranged between the first and third sodium chamber for cooling the core support.

Abstract: An emergency feedwater system for the steam generators of a pressurized water nuclear reactor has two separately located subsystems, each subsystem supplying water to at least one steam generator when activated, where each subsystem contains an emergency feedwater supply tank, and a pair of emergency feedwater lines leading from the tank which communicate with the inlet line of a steam generator. An electrical operated motor driven pump is located in one of said pair of emergency feedwater lines and a steam turbine driven pump in the other of said pair, with a cavitating venturi provided in an emergency feedwater line between the pump and inlet line of the steam generator. The system is adaptable for use with two, three, and four loop pressurized water reactor systems.

Abstract: The invention relates to a standby or emergency cooling device for the core of a pressurized water reactor.A high pressure reservoir filled with boric acid solution is permanently connected to the reactor vessel via two pipes, preferably arriving at the same level. However, one of these pipes is extended within the vessel by a downwardly directed bend in order to issue below the other pipe. These two pipes have horizontal portions which are intended to prevent any natural circulation under normal operating conditions.Application to the improvement of the safety of pressurized water nuclear reactors.

Abstract: Nuclear reactor cooled by a liquid metal comprising a main vessel sealed by a sealing slab, said vessel containing liquid metal overhung by a covering of inert gas, the reactor core being placed in a primary vessel integrated into the main vessel in such a way that the primary vessel at the outlet from the core constitutes a hot collector for the liquid metal and the space defined between the main vessel and the primary vessel constitutes a cold collector for the liquid metal, at least one heat exchanger supplied by the liquid metal contained in the hot collector under the action of at least primary pumps located in the main vessel, wherein the exchanger is positioned outside the main vessel at a level above that of the sealing slab and communicates with the hot collector by at least one intake pipe and with the cold collector by at least one outlet pipe, it being possible to empty the exchanger and pipes into the main vessel by gravity, the upper part of each exchanger being filled with the inert gas coverin

Abstract: A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. The reactor vessel, reactor coolant supply loops, emergency water storage tank and pump modules are arranged in separate compartments in the containment structure to control the flow of spilled coolant. An integrated containment spray system and normal cooldown system are also described.

Abstract: An emergency cooling device for a water-cooled nuclear reactor comprises a heat exchanger integrated into a reactor vessel and submerged in the reactor water. One of the ends of the circuit of this exchanger is connected by means of a three-way valve to the upstream valve, while its other end is connected across a downstream valve and a second untreated water-operated cooling exchanger to the cold water tank.

Abstract: Device for the post-accident cooling of the confinement enclosure of a pressurized water nuclear reactor, wherein it comprises in series a turbine supplied by the hot, humid air contained in the enclosure, a condenser in which the air from the turbine is dried and cooled by thermal contact with an external cooling fluid, and a compressor actuated by the turbine, and which returns the dried air into the enclosure.

Abstract: A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure.If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel.

Abstract: Installation for cooling a water-filled fuel element storage well, including a plurality of cooling circuits having drain lines with connections to the well, pumps connected to the drain lines, heat exchangers connected to the pumps and feed lines being connected to the heat exchangers and having connections to the well, one of the connections of the drain lines to the well being disposed at a level which is at least as high as the other of the drain line and feed line connections, and a downwardly directed pipe disposed in the storage well being connected to the one of the drain line connections and having a lower opening formed therein at a lower level than the drain line and feed line connections.

Abstract: A pressurized water nuclear reactor comprises a reactor vessel arranged in a pool made in the form of a pressure vessel, the pool being filled with a strongly neutron-absorbing liquid, for example borated water. The reactor vessel, in addition to being connected to a primary system, is provided with a lower shutdown opening, arranged below the reactor core, for pool liquid and with an upper shutdown opening, arranged above the core, for pool liquid.

Abstract: A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors.

Abstract: A standpipe runs up from a liquid storage vessel located between the safety vessel of a nuclear reactor and the safety enclosure around it and connected by a pressure equalization line to the interior space of the safety vessel. Undesired pressure increases in the interior of the safety vessel are absorbed by pushing water out of the storage vessel up the standpipe to a discharge above the safety vessel which is still within the safety enclosure. With sufficient pressure, the liquid, which may be conveniently water, flows from the standpipe to an overflow collar at the top of the safety vessel from which it runs down over the surface of that vessel to cool it. Water that evaporates is collected on the interior wall of the safety enclosure and also percolates down to the base of the safety enclosure, from which it is pumped back to the overflow collar, from which it can flow back into the pressure vessel when the pressure inside the safety vessel of the reactor goes back to normal.

Abstract: The invention relates to a fast neutron nuclear reactor equipped with residual power removal devices.Each device comprises an evaporator incorporating a bundle of tubes in glove finger-like form immersed in the liquid metal contained in the reactor vessel, an adiabatic collector constituted by a pipe traversing the slab sealing the reactor vessel and a condenser in which the heat transfer fluid, such as mercury, contained in said device is condensed by heat exchange with the atmospheric air sucked in through the chimney or flue.Application to fast neutron nuclear reactors.

Abstract: A method of cooling the primary circuit of a pressurized water nuclear reactor in which, after a first stage, during which water is injected into the steam generators and the steam produced is discharged, cooling is continued during a second stage by water-water heat exchange. In this second stage, the water is made to circulate at the secondary side of at least one steam generator (2) in countercurrent to the primary water, over at least part of its path. The secondary water heated by the primary water is cooled outside the containment enclosure (1) of the nuclear reactor in at least one heat exchanger (31) using raw cooling water. The secondary water is recycled in the steam generator (2). The invention is applicable to the effecting an maintaining of cold shutdown of pressurized water nuclear reactors.

Abstract: An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe.

Abstract: The invention relates to a nuclear reactor cooled by a liquid metal contained in a vessel obturated by upper closing means, said latter essentially being constituted by a slab and rotating plugs. The upper closing means comprise at least one circuit for circulation of a cooling gas or liquid, this circuit being provided with means for maintaining the closing means at a homogeneous temperature at least equal to 100.degree. C. and preferably close to 120.degree. C. The invention is applicable to the production of less cumbersome and less expensive fast neutron nuclear reactors.

Abstract: Process and device for emergency cooling of a nuclear reactor comprising a steam generator (2) heated by a primary circuit and cooled by a secondary circuit producing steam, and in which in case of an emergency shutdown of the reactor, the emergency feed (50) of the secondary circuit is provided from an emergency reservoir (5), by an auxiliary pump (51) driven by steam derived at the output from the generator. At least part of the steam continuing to be produced during the cooling process is condensed in an air condenser (6) and the condensed water is recycled to the emergency reservoir (5). The power for the air condenser is supplied from another withdrawn portion (70) of steam.

Abstract: An improved filling for emergency cooling towers of nuclear power plants, of the type made of a grid of tubular rods in plastic material, such as PVC, has an anchoring system by which it can withstand atmospheric depressions or overpressures up to 700 Kg/m.sup.2, as required. The tubes of plastics, arranged in pairs of staggered, spaced parallel rows, are inserted to pass through two opposite walls of at least one elongated square section, perpendicular thereto and preferably of the same plastic material. The distance between the two rows is slightly less than the diameter of an auxiliary stiffering tube which is inserted between said two section walls and the two rows of tubes, transversely thereto. For each layer of filling, a particular system of fixing said square sections to the concrete structure of the tower is described to meet also the aseismatic requirements.

Abstract: In this process, the decay heat of radioactive substances is carried away by circulating liquid coolant. Some of the liquid coolant is vaporized by the decay heat. The circulation of liquid in the circuit is driven by pressure from the vapor. After exceeding a static pressure head corresponding to the pressure drop in the circuit, the vapor is separated from the liquid and condensed, and the condensate is combined with the liquid returning for repeated partial vaporization.

Abstract: The invention relates more particularly to a pressurized water nuclear reactor in which the heat exchange between the primary and secondary circuits takes place by means of steam generators.A reserve tank is placed in the secondary circuit parallel to the steam generator and inside or outside the latter. It is pressurized directly by the secondary fluid in the gaseous phase leaving the generator, so as to form a reserve increasing the inertia of the generator in the case of closing the steam and supply valves without significantly increasing it in the case of a fracture of the steam piping.Application to pressurized water nuclear reactors.

Abstract: The system according to the invention comprises a first room which accommodates a reactor plant and an active-type sprinkler means to condense steam released by the ebullient coolant after an accident. A loss of coolant may occur at any point of the reactor plant; that notwithstanding, the rising pressure in said first room drives the air out of that room through holes provided in the walls of the room in immediate proximity to the floor and spaced at a maximum distance from one another over the perimeter of the walls. The air is forced into an intermediate chamber wherefrom it is directed to a second room through channels and a basin-type condenser arranged in a second room. The length of the channels is selected so as to form a water seal to prevent the backflow of air driven from the first room to the second room and thus produce rarefaction in the first room.

Abstract: Integrated nuclear reactor cooled by a liquid metal and incorporating a main vessel sealed in its upper part by a slab, an inner vessel containing the core, the latter resting on a system for the positioning and supply of the core with liquid metal and which is called the support, the latter itself resting on a supporting structure bearing on the bottom of the main vessel of the reactor, wherein it comprises an inner baffle cladding the side wall and bottom of the main vessel and defining with the latter an intermediate space filled with the liquid metal, tubes for supplying a liquid metal to the intermediate space below the bottom of the main vessel and tubes for returning said liquid metal to an auxiliary exchanger in order to remove heat from the intermediate space.

Abstract: Cooling system for auxiliary systems of a nuclear installation for heat removal from heat exchangers, the heat exchangers being connected on the primary side thereof to lines which may contain radioactive liquids or gases, being disposed in a secured area of the nuclear installation, and having connections on the secondary side thereof for cooling liquid lines, including an outgoing line for the cooling liquid connected to the connections on the secondary side of the heat exchangers, a dry cooling tower having cooling elements connected to the outgoing line, a return line for the cooling liquid connected to the cooling elements, a refrigeration loop having a supplemental heat exchanger with the primary side thereof connected in the return line, a bypass line connected from the outgoing to the return line parallel to the cooling elements and supplemental heat exchanger, and a control valve connected in the bypass line.

Abstract: A sodium-cooled fast reactor comprises at least one auxiliary heat exchanger constituted by a plurality of vertical heat-exchange modules each provided with a coolant circuit. A vertical cylindrical passage for each auxiliary heat exchanger extends through the reactor vault roof. The passage diameter is larger than the diameter of a heat-exchange module but smaller than the overall dimension of the auxiliary heat exchanger in horizontal cross-section. Each passage is closed by a shield plug after the modules have been placed in position, the modules being disposed around an axial extension of each shield plug.

Abstract: An apparatus for cooling molten material resulting from a nuclear reactor core meltdown is disclosed. The apparatus includes a basin positioned under the reactor which is protected against excessive heat by a star-like array of heat pipes whose evaporator sections are disposed above the pan and whose condenser sections are disposed in a heat sink exterior to the containment building of the reactor. Additionally, the vertical walls of the reactor vessel chamber are similarly protected by an array of heat pipes similarly arranged and provided to intercept the radient energy of the molten core material.

Abstract: A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Abstract: The cavity within a pressure vessel for a gas-cooled nuclear reactor is divided into chambers by an intermediate support floor, with the reactor core thereabove and heat-exchangers therebelow. Separate heat-exchange units are connected to separate circulators which discharge the repressurized primary gas coolant through an annular passageway along the inner surface of the pressure vessel cavity. The heat extraction system includes a separate secondary gas coolant circuit connected to each heat-exchange unit, which coolant may be heated to about 850.degree. C. at which it is suitable to provide process heat. Emergency cooling is effected by a heat dump which is connected into each secondary circuit to promptly provide low-temperature secondary gaseous coolant which then indirectly removes the reactor core heat.