Abstract: A reactor core with each fuel rod surrounded by an individual cylindrical channel is disclosed. This individual channel on each fuel rod provides thermal hydraulic and heat transfer advantages to enable all fuel rods within the fuel bundle to uniformly approach their own thermal limits. The preferred fuel rod pitch is a triangular pitch between the individual fuel rods as they are discretely surrounded by their own channel. The new triangular geometry provides for more uniform (flat) power distributions within all fuel rods--and hence all groups of fuel rods. Bypass flow is introduced uniformly between the fuel rod channels, rather than heterogeneously in the channel gap and water rods as in present BWR fuel designs. Individual fuel rod channels can be orificed differently, as required, to match inlet flow to fuel rod power output to maintain uniformity between all fuel rods as they approach their respective thermal limits.

Abstract: A satellite heat removal means can be embodied as an original nuclear system feature but is especially adapted to be retrofitted to an existing nuclear reactor system to serve optionally to supplement heat removal from the system nuclear reactor containment upon happening of a LOCA, and to assume all system containment drywell venting in the event reactor core meltdown results in breach of the containment floor structure separating the containment drywell and wetwell space which breach would deprive the containment of a space to which a non-condensable fraction of LOCA generated heated fluid in the containment could be vented, cooled and stored. The satellite heat removal means includes a structural external of but preferably situated alongside the nuclear reactor containment. A heat exchanger surrounded by a pool of cooling water is located in an upper chamber of the structure while a pool of water is present in a lower chamber of the structure.

Abstract: There is disclosed a nuclear reactor installation. An accumulator-type emergency core cooling system, a gravity-driven core cooling system and an equalizing system for submerging a reactor core are provided within a primary containment vessel containing a reactor pressure vessel in which the reactor core is disposed. These cooling systems are automatically operated sequentially in accordance with the pressure in the reactor pressure vessel without the need for any particular powered source. The primary containment vessel is made of steel, and the interior of this containment vessel is divided into a space containing the reactor pressure vessel and a space containing an operation floor in such a manner that the two spaces are isolated from each other.

Abstract: The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Abstract: In combination with a nuclear power generating facility including a composite fuel pool including a reactor cavity and a spent fuel pool fluidly connectable to the reactor cavity, the composite fuel pool at least partially containing a primary fluid, a nuclear reactor vessel positioned in the reactor cavity, a residual heat removal system installed in the facility and fluidly connectable to the reactor vessel, and a spent fuel pool cooling system installed in the facility and fluidly connectable to the spent fuel pool, a temporary cooling system is provided, comprising a primary heat exchange system including a primary heat exchanger for transferring heat from a primary fluid to a secondary cooling fluid. The primary heat exchanger is temporarily locatable in the facility, and is temporarily fluidly connected to the composite fuel pool.

Abstract: A water storage tank in the coolant water loop of a nuclear reactor contains a tubular heat exchanger. The heat exchanger has tubesheets mounted to the tank connections so that the tubesheets and tubes may be readily inspected and repaired. Preferably, the tubes extend from the tubesheets on a square pitch and then on a rectangular pitch therebetween. Also, the heat exchanger is supported by a frame so that the tank wall is not required to support all of its weight.

Abstract: A sodium cooled fast ractor comprises a reactor vessel in which a liquid metal coolant is accommodated, a core disposed substantially a lower central portion of the reactor vessel in an installed state, a core support structure secured to the reactor vessel for supporting the core, the core support structure dividing an interior of the reactor vessel into a high-pressure plenum below the core and a low-pressure plenum above the high-pressure plenum, a circulation pump unit for applying a discharge pressure to the liquid metal coolant and circulating the same, and an intermediate heat exchanger for performing a heat exchanging operation of the coolant in the reactor vessel. The circulation pump unit is composed of an electromagnetic circulation pump provided with a discharge port and a closed gas space, which is filled up with a closed gas, defined above and communicated with the discharge port.

Abstract: If an emergency arises in gas-cooled nuclear reactor plants having a heat exchanger for removing afterheat due to failure of a blower, the afterheat can be only inadequately removed because of the many deflections and restrictions in a primary circulation loop. In order to achieve adequate removal of afterheat solely on the basis of natural draught, a heat exchanger is located above a top reflector and reaches across a part of coolant gas bores in the top reflector. During normal operation of the nuclear reactor plant, there is a flow of cold gas through the heat exchanger in the direction toward the reactor core and in the event of an emergency, there is a flow of hot gas in a direction out of the center of the core of the reactor. A natural draught coolant circulation is established with the return of the cooled gas through the coolant gas bores in the top reflector.

Abstract: A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel.

Abstract: A control room 10 for a nuclear plant is disclosed. In the control room, objects 12, 20, 22, 26, 30 are no less than four inches from walls 10.2. A ceiling 32 contains cooling fins 35 that extend downwards toward the floor from metal plates 34. A concrete slab 33 is poured over the plates. Studs 36 are welded to the plates and are encased in the concrete.

Abstract: Groundwater running below the surface is intercepted with a cutoff wall of a predetermined height and length to form an underground dam, and a nuclear power generation system is built below said underground dam. The water pooled in the underground dam is utilized as the secondary cooling water for the reactor of the power system.

Abstract: A nuclear reactor system including a main heat transport path around which a first coolant is pumped by main circulation pumps to transport heat from the reactor core to a steam generator. A heat exchanger is located in the main heat transport path after the outlet from the steam generator, the secondary side of that heat exchanger being in a decay heat removal loop which contains a second liquid coolant. A vapor separator in the decay heat removal loop is connected to an outlet of the heat exchanger with the vapor separator's outlet being connected to an inlet of a further heat exchanger located in a large tank of water which forms a heat sink. The further heat exchanger's outlet is connected to the heat exchanger's inlet forming a closed loop. The further heat exchanger is located at a higher elevation than the heat exchanger whereby a natural convection flow can occur in the decay heat removal path.

Abstract: A conventional low pressure coolant injection system for boiling water reactors is provided. With the modification, the cross tie conduits and associated valves remain open between two LPCI divisions. On the occasion of an LOCA, the LPCI pumps are activated and injection valves for each of the LPCI divisions are opened to commence coolant injection in the recirculation loops in simultaneous fashion. However, the rate of flow of water coolant within each injection system is controlled by a hydraulic resistance, which is selected to achieve reactor core cooling within requisite time and in requisite quantities from one injection path. Thus, even though coolant water may flow through a rupture within one recirculation loops, sufficient water will be injected in the other loop to achieve core cooling.

Abstract: A nuclear energy plant housing a boiling water reactor utilizes an isolation condensers with nearly horizontal condenser tubes for both isolation condenser (IC) mode and passive containment cooling system (PCCS) mode, which is entered in response to a loss-of-coolant accident (LOCA). These tubes extend between a cylindrical distributor and a cylindrical collector. In either mode, the reactor vessel is coupled to the cylindrical distributor. Steam reaching the condenser is condensed to water, which flows back to the vessel, providing a cooling effect. In PCCS mode, gas exiting the tubes is trapped and diverted into a wet well. This diversion path is not avialable in IC mode. As a result water exits the tubes more slowly in IC modes in PCCS mode. The water remaining in the tubes during IC mode renders the condenser less efficient. This lower efficiency partially offsets the greater heat exchange in IC mode due to higher temperature differentials and the relatively absence of noncondensable gases.

Abstract: A boiling water nuclear water plant includes a reactor core isolation cooling (RCIC) system in which a turbine used to pump feedwater to the reactor also drives a generator. The generator is used to drive RCIC components, such as a room cooler and control electronics, during station blackout.

Abstract: A ventilating system for an enclosure in a nuclear power plant is provided. The enclosure houses an emergency feedwater pump and its driver steam turbine. An eductor mounted in the enclosure is supplied with steam exhausting from the driver steam turbine. The steam jet formed by the eductor draw hot air out of the enclosure, thereby preventing the excessive build-up of heat in the enclousre. An air inlet formed in another wall of the enclosure supplies the enclosure with cool ambient air. The eductor and air inlet are situated to ensure the incoming air flows over the pump, thereby cooling it. A duct transports the steam/air mixture discharged by the eductor to the atmosphere.

Abstract: A unique and optimum nuclear steam supply system operating configuration for integration of a chemical decontamination system is disclosed. The chemical decontamination system is connected to, and returns to, the residual heat removal system downstream of a residual heat removal heat exchanger, thereby utilizing the residual heat removal system to control the temperature of process fluids entering the decontamination system. A reactor coolant pump or pumps generates heat for the chemical processes as needed and a nitrogen blanket within the primary system pressurizer is utilized for system pressure control.

Abstract: A method of avoiding localized hydrogen build-ups in the atmospheres of safety tanks of reactors, especially light water reactors, is provided. After a break down that is accompanied by loss of coolant has occurred, the cooling water temperature in the sump of the safety tank is adjusted in the phase of the long term cooling to a temperature that is high than the air temperature in the dome of the safety tank, with this temperature adjustment being effected in such a way that an adequate intermixing of the atmosphere in said safety tank is effective in a convective manner.

Abstract: A multiple liquid standby safety injection system for nuclear fission reactor plants comprising means for injecting supplemental coolant water into the nuclear reactor pressure vessel to cool the fuel core and means for injecting a water solution of a neutron absorbing compound into the nuclear reactor pressure vessel about the fuel core to diminish the fission reaction. The coolant water and solution of neutron absorbent each comprise individual systems and are conveyed from their respective supply container by means of pressurized propelling gas. The individual standby safety injection systems for coolant water and solution are integrated with means for transferring propelling gas from one supply container to the other to enhance the source and available volume of liquid propelling gas for either system by drawing from the other.

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: 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 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 pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

Abstract: A suppresion chamber in a nuclear reactor containment includes a heat exchanger disposed in a gas space above water in the suppression chamber. A gravity-driven pool contains a supply of make-up water that is gravity fed to the heat exchanger through a conventional level-maintaining valve such as a float valve. A top header in the heat exchanger includes a free surface area for permitting separation of steam from water, thereby permitting venting of vapor only, while retaining the liquid coolant in the heat exchanger. A downcomer tube permits return of excess water to a lower location for further use in the heat exchanger. The heat exchanger is normally sealed, whereby internal surfaces in the heat exchanger require only a small amount of low-volatility corrosion inhibitors to prevent corrosion on internal surfaces thereof. Locating the heat exchanger in the gas space in the suppression chamber reduces the amount of corrosion on its external surfaces.

Abstract: A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

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: Pressurized water reactor with a primary circuit including therein a reactor pressure vessel, a steam generator and a main coolant pump, and with an auxiliary system having high pressure pumps for feeding water into the primary circuit, including a line extending from an upper side of the pressure vessel and having at least one shut-off valve therein, the line connecting the reactor pressure vessel and a part of the auxiliary system wherein a lower pressure prevails than in the reactor pressure vessel.

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: The apparatus for simulation of the operation of a pressurized water reactor comprises a reactor vessel with a heater, a steam generating vessel divided into a primary portion and a secondary portion. The reactor vessel and the primary portion of the steam generating vessel are connected in a primary feed loop equipped with a coolant pump, while the secondary portion of the steam generating vessel and the condenser unit are joined together in a secondary feed loop provided with a feed pump. With such an apparatus a steam generating vessel heating pipe break or fracture can be simulated when the reactor vessel and a primary part of an additional steam generating unit are joined in another primary feed loop parallel to the primary feed loop having the cooling pump and a secondary part of the additional steam generating unit is connected with the aid of at least two branch pipes to the secondary feed loop.

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: 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 high-temperature gas cooled nuclear reactor system comprises a containment building, a concrete reactor pressure vessel inside the containment building, and a safety relief valve connected to the concrete reactor pressure vessel. The spring of the safety valve consists of a material with a spring constant decreasing as temperature rises. A heat exchanger is provided in close proximity to cool the spring of the safety valve which is subject to the heat of the reactor coolant escaping when the safety valve is open. The heat exchanger of the safety valve is connected to a liner cooling system of the concrete reactor pressure vessel.

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: An entirely passive auxiliary core cooling system for a liquid-metal reactor wherein a path for natural circulation through a radial plenum and radially outermost core assemblies to the heat generating inner core assemblies is provided, the flow being cooled by a totally passive heat exchanger.

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 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: 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 nuclear reactor plant housed in a steel pressure vessel with a small high temperature reactor, together with a steam generator arranged over said reactor and with at least two circulating blowers connected in parallel with each other and following the steam generator in line and is provided with a decay heat removal system located in the primary loop. This system is arranged above the steam generator and a shut-off device is provided between the two component. The shut-off device, which may actuated both passively--by difference pressure--and actively, by means of a drive, is designed so that in normal operation no hot gas may enter the decay heat removal system. In the decay heat removal mode the hot gas is conducted into the decay heat removal system and the steam generator is closed off the hot gas. Simultaneously, the cold gas coming from the decay heat removal system is returned through the steam generator or a gas conduit parallel to it, to the reactor core.

Abstract: A pressurized nuclear water reactor has a substantially cylindrical flow liner with a cylindrical wall section and bottom and an open top. A barrel forms a riser chamber that contains the core in the flow liner. A pressure vessel contains the cylindrical flow liner to form a second annular chamber therebetween that contains a supplementary liquid coolant, with insulation means to provide a major portion of the supplementary liquid coolant at a first temperature and a minor portion thereof at a second higher temperature. Upon depressurization in the vessel, fluid communication means enable injection of supplementary liquid coolant from the second annular chamber into the core upon flashing of a minor portion to vapor. A further pool of water outside the pressure vessel, and insulation on the wall, maintain the desired temperature in the supplementary liquid coolant supply. Injection or removal of borated solution, as a chemistry control solution, into or from the supplementary liquid coolant is provided.

Abstract: The invention relates to a heat exchanger incorporating emergency cooling means, as well as to a fast neutron nuclear reactor incorporating such an exchanger.The emergency cooling means comprise an inlet pipe issuing into the secondary fluid inlet chamber and an outlet pipe issuing into the secondary fluid outlet chamber. The inlet pipe issues into a distributing ramp positioned below the lower tue plate and the outlet pipe issues into a collecting ramp positioned above the upper tube plate. The two ramps are positioned below the intake and discharge pipes for the secondary fluid.

Abstract: A process for the controlled discharge from a reactor containment structure of a gas cooled nuclear power plant and the installation for achieving this process include parallel discharge circuits from a reactor containment structure to a discharge stack. The flow of discharge from the reactor containment structure normally flows in a primary discharge circuit and may be directed wholly or partially to the parallel secondary discharge circuit. The secondary discharge circuit contains means for reduction of the temperature of the discharge, means for deposition of particulate fission products and they also contain means for recombination of combustible gases and means for filtration of the discharge. Nuclear power plant installations and processes for controlling the discharge from a reactor containment structure in this fashion permit safe discharge of naturally occurring leakage for gas cooled nuclear reactors, as well as leakage occurring in the event of reactor failure.

Abstract: In a nuclear reactor of the type in which the vault has a roof in which inner and outer rotatable shields are mounted, coolant is supplied to the roof and shields via duct penetrations which extend in the plane of rotation and can be brought into alignment and connected together by slidable connectors when relative rotation between the shields and roof is not required. When such rotation is to be effected, the connectors can each be retracted into a respective duct so that rotation is not obstructed thereby.

Abstract: The nuclear reactor installation is provided with a double-wall pressure vessel which is surrounded by a heat sink. A flood tank is provided outside the vessel in order to supply air as a heat-insulating agent to the chamber defined by the pressure chamber wall during normal operation and to supply water as a heat-conductive agent in response to a disturbance in reactor core cooling.The cover on the pressure vessel is also provided with a chamber to which air or water can be delivered. A closure member which can be activated by fusible element is incorporated in the cover to permit flooding of the cover chamber in response to an excessive heat within the pressure vessel.

Abstract: A nuclear reactor plant with a small high temperature reactor, principal heat exchangers located above the small high temperature reactor and housed in a steel pressure vessel, and a plurality of decay heat exchangers also located in the steel pressure vessel and connected on the cooling water side with an external recooling heat exchanger each, in a geodetically higher location. The object of the invention is to obtain a high degree of availability of the decay heat removal installations. This object is attained by the specific layout and connection of the components of the plant, which permits the operation of the decay heat exchangers both on the primary gas and the water side by natural convection only. The decay heat exchangers and their secondary circulation loops are active during power operations, so that potential failures and leaks are detected immediately. No special actuating measures are necessary to activate the decay heat removal operation.

Abstract: The present invention provides a passive safety device for dumping steam from the steam generator of a nuclear power reactor in case of emergencies. The device comprises a steam ejector immersed in a storage tank of coolant water, a conduit connecting the steam ejector to the steam line from the steam generator, a heat exchanger having its inlet connected to the outlet of the steam ejector and the outlet to the storage tank, and a coolant pool in which the heat exchanger is immersed. Optionally, the output from the steam ejector may be connected to the feedwater line to the steam generator as an emergency feedwater supply.

Abstract: In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Abstract: The invention concerns a safety cooling installation for the water reactor of a nuclear power station, the installation notably incorporating a reservoir for storing a cooling liquid, the said reservoir being located outside the containment which encloses the reactor circuit, the installation being provided with aspiration and discharging means arranged to aspirate the liquid from the reservoir and to discharge it into the circuit, there being provided a passive, maintenance free means to recover and conduct the liquid and containment water, the said means being arranged to recover the liquid and the water from the reactor in the lower part of the containment by non gravitational flow in the event of a break in the reactor's circuit.

Abstract: A heat-generating member (4) is arranged in a water-filled pressure vessel (1) which is provided with a pressure relief valve or the like. The water of the pressure vessel can be partly evaporated, thereby acting as a heat sink for the generated heat. The walls of an outer vessel (6) surround at least a lower part of the pressure vessel (1) in such a way that a closed, relatively small auxiliary space (7) is formed between the two vessels. The auxiliary space (7) communicates via at least one tube (8) with an open evaporation pool (9), which is arranged above the cover (2) of the pressure vessel (1). A tube coil (10), disposed in an upper part of the pressure vessel, is connected by both ends to the evaporation pool (9).

Abstract: A passive depressurization system for use during cold shutdown of a pressurized water reactor receives saturated steam from the reactor's pressurizer, condenses the steam to water, and returns the water to a nozzle within the pressurizer in order to provide a cooling spray which condenses steam within the pressurizer. The depressurization system may include coiled tubing, which is connected to input and output ports of the depressurization system by redundant valves, or a condenser formed by a housing having air vents provided by vent elements through the housing.