Abstract: The present disclosure provides a stopped cooling system including: a steam line connecting portion connected to a steam line so as to receive cooling water through the steam line connected to an outlet of a steam generator; a stopped cooling heat exchanger for receiving cooling water that enters the stopped cooling system through the steam line connecting portion, and discharging same through a passage of the heat exchanger; a stopped cooling pump activated to perform stopped cooling of the nuclear reactor upon normal stoppage of the nuclear reactor after primary cooling of the nuclear reactor cooling system or when an accident occurs, and for forming a circulating flow of cooling water that circulates between the steam generator and the stopped cooling heat exchanger; and a water supplying pipe connecting portion connected to the heat exchanger passage and a water supplying pipe, which is connected to the inlet of the steam generator, so as to supply the cooling water cooled in the stopped cooling heat exch

Abstract: A light water reactor for generating power that utilizes circulation of a primary coolant at saturation pressure to cool a nuclear core and transfer heat from the core to a secondary coolant through one or more heat exchangers of a condensing steam generator. The secondary coolant, once heated can drive power generation equipment, such as steam turbines or otherwise, before being condensed and returned to the one or more heat exchangers.

Abstract: A nuclear reactor includes a nuclear reactor core disposed in a pressure vessel and immersed in primary coolant water at an operating pressure higher than atmospheric pressure. A containment structure contains the nuclear reactor. A reactor coolant inventory and purification system (RCI) is connected with the pressure vessel by make-up and letdown lines. The RCI includes a high pressure heat exchanger configured to operate responsive to a safety event at the operating pressure to remove heat from the primary coolant water in the pressure vessel. An auxiliary condenser located outside containment also removes heat. The RCI also includes a pump configured to inject make up water into the pressure vessel via the make-up line against the operating pressure. An emergency core cooling system (ECC) operates to depressurize the nuclear reactor only if the RCI and auxiliary condenser are unable to manage the safety event.

Abstract: A passive high-pressure safety injection system includes a compressor which generates high-temperature and high-pressure steam, a cooling water supply tank which supplies cooling water using the compressed steam, a nuclear reactor which receives the cooling water so that the nuclear reactor is maintained in a cooled state, and an internal circulation prevention structure which is provided in the cooling water supply tank and prevents the cooling water from circulating in the cooling water supply tank.

Abstract: This device comprises a spray assembly (40) for spraying fluid into the chamber, and a fluid feed pipe (42) intended to feed fluid to the spray assembly (40). It comprises a fluid-distributing intermediate receptacle (46) positioned between the feed pipe (42) and the spray assembly (40), the intermediate receptacle (46) being connected upstream to the feed pipe (42) and comprising a fluid-evacuating sidewall (52) delimiting through orifices (60) connected to the spray assembly (40). It comprises at least one pipe (44A to 44D) for evacuating fluid towards the chamber (19) and projecting into the intermediate receptacle (46) opposite the sidewall (52).

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

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

Abstract: This device comprises a spray assembly (40) for spraying fluid into the chamber, and a fluid feed pipe (42) intended to feed fluid to the spray assembly (40). It comprises a fluid-distributing intermediate receptacle (46) positioned between the feed pipe (42) and the spray assembly (40), the intermediate receptacle (46) being connected upstream to the feed pipe (42) and comprising a fluid-evacuating sidewall (52) delimiting through orifices (60) connected to the spray assembly (40). It comprises at least one pipe (44A to 44D) for evacuating fluid towards the chamber (19) and projecting into the intermediate receptacle (46) opposite the sidewall (52).

Abstract: A feeding system for an absorber liquid containing a neutron poison, in particular for a quick shut-down of a nuclear reactor, has a storage container for the absorber liquid and is configured for high operational reliability with simple construction. In particular, a chemical decomposition of the absorber liquid or corrosion of the container wall of the storage container is to be excluded. For this purpose, the storage container is connected to a pressure container via an overflow line, wherein the pressure container is filled with a motive fluid.

Abstract: A feeding system for an absorber liquid containing a neutron poison, in particular for a quick shut-down of a nuclear reactor, has a storage container for the absorber liquid and is configured for high operational reliability with simple construction. In particular, a chemical decomposition of the absorber liquid or corrosion of the container wall of the storage container is to be excluded. For this purpose, the storage container is connected to a pressure container via an overflow line, wherein the pressure container is filled with a motive fluid.

Abstract: A pressurizer is for a pressurized water nuclear power plant and it includes an upper cap provided with a tube; an end piece connected to the tube using a weld; and a sleeve protecting the weld, disposed inside the tube. The protective sleeve is mounted in a removable manner, such that the thermal sleeve is removed.

Abstract: A pressurizer whose heater sheaths are conditioned to reduce the residual stresses resulting from cold working during manufacture. After material conditioning, the heater sheath undergoes a surface conditioning treatment to add outer surface compressive stresses.

Abstract: In order, in a pressure accumulator, to generate a required operating pressure of a fluid situated in the pressure accumulator, a heater device is provided, disposed in an upper region of the pressure accumulator. The device generates a vapor cushion and the required operating pressure. Where necessary, when the fluid flows out, the pressure in the pressure accumulator is automatically reduced after the level has fallen below a defined filling level, as a result of the vapor condensing at a cold insulating device with a high heat-accumulation capacity. The pressure accumulator is preferably used as an emergency shut-down tank or as an emergency accumulator in a nuclear power plant.

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

Abstract: A pressurizer includes a casing in which at least one spray line ends. The spray line is guided through a wall in a lower region of the casing and is disposed in such a way as to run upward inside the casing. The spray line ends at its geodetically highest point.

Abstract: A nuclear reactor coolant system includes a primary coolant circuit connected to a secondary coolant circuit. A desired amount of coolant is bled from the primary circuit to the secondary circuit for purification and controlling chemical composition of the coolant. Variable speed charging pumps are provided in the secondary circuit to pump coolant back into the primary circuit at variable pressures and flow rates. A pressurizer level control system is also disclosed for controlling pump speeds.

Abstract: For nozzle replacement, repair and initial installation in ASME pressure vessels, disclosed are nozzles, nozzle assemblies and nozzle repair assemblies, all of which are mechanically attached and mechanically sealed to the vessel, i.e., without any welding to the vessel, and in most embodiments without any welding at all. For nozzle replacement, a part or the entire existing nozzle is removed and a partial or full replacement nozzle or nozzle assembly is mechanically attached and mechanically sealed to the vessel. For nozzle repair, the existing nozzle within the bore is not removed, and a mechanical seal is provided for the existing nozzle which may also be welded or mechanically attached to the repair assembly.

Abstract: The recovery device includes a first tank containing a volume of water open in the inner space of the safety containment and a closed second tank isolated from the inner space of the safety containment. The second tank is smaller than the first tank, and is at least partly immersed in the water contained in the first tank. The inner volume of the immersed part of the second tank is placed in communication with the volume of water contained in the first tank by strainers. A discharge pipe is connected at one of its ends to a discharge point of the coolant circuit of the nuclear reactor and includes an end part opposite the discharge point opening into the interior of the second tank.

Abstract: The invention disclosed herein provides nozzles for vessels such as a pressurizer vessel, reactor vessel or piping in a nuclear power facility, and methods for attaching and sealing the nozzles to the vessels without any welding at all, i.e., without using a structural or seal weld. The nozzles are clamped or bolted to the vessel or attached by an interference fit, and a mechanical seal is obtained using gasket material and/or contacting, preferably polished, metal surfaces. Some embodiments of clamping the nozzle to the vessel comprise tightening a nut on the threaded end of the nozzle projecting from the bore in the vessel against the vessel, and structure which prevents the nozzle from being withdrawn through the bore, thereby clamping the nozzle against the vessel. Another embodiment of clamping a nozzle to the vessel comprises a sleeve threaded in the bore and forcing an interfering portion of the nozzle body into engagement with the bore.

Abstract: An automated method for maintaining pressure within a nuclear power plant primary loop during either startup or shutdown, the method comprises the steps of partially filling a portion of a pressurizer vessel, in fluid communication with the primary loop, with a liquid for maintaining pressure in the primary loop; circulating a primary coolant through the primary loop; automatically injecting an inert gas by a first automated device, operatively connected to the pressurizer, into the pressurizer vessel when the pressure in the pressurizer vessel is less than a first predetermined pressure; and automatically venting the gas by a second automated device, operatively connected to the pressurizer, from the pressurizer vessel when the pressure in the pressurizer vessel is greater than a second predetermined pressure.

Abstract: An equalizing reservoir configuration for fluid circuits with volume and/or pressure fluctuations, especially for cooling circuits in nuclear power stations, includes an equalizing reservoir communicating with a fluid circuit through a fluid line connected to a base of the equalizing reservoir, so that a lower fluid chamber and an upper gas buffer chamber form in the equalizing reservoir. An upper end of an equalizing riser opens into the gas buffer chamber and runs out into a U-tube containing a water/air seal and being disposed beneath the equalizing reservoir. This provides a slim construction for narrow spaces and an increase in the air volume of the equalizing reservoir.

Abstract: A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90.degree. intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity.

Abstract: A method for replacing a nozzle in a pressure vessel. The existing nozzle is cut approximately at the inside surface of the pressure vessel wall. The portion of the existing nozzle that extends beyond the exterior of the pressure vessel wall is removed. A weld pad is deposited on the exterior of the pressure vessel wall around the nozzle bore. The remainder of the existing nozzle is removed. A corrosion resistant thermal spray coating is applied to the nozzle bore. A replacement nozzle is installed in the nozzle bore and seal welded to the weld pad on the exterior of the pressure vessel wall.

Abstract: A replacement nozzle and method for replacing a nozzle in a pressure vessel. The original nozzle is removed and a weld preparation is machined into the original weld. The vacated nozzle hole is partially drilled out from the interior of the vessel. A weld land sleeve is received in the drilled out portion of the nozzle hole and a replacement nozzle is received in the weld land sleeve. The sleeve and replacement nozzle are welded in place on the inside of the vessel wall.

Abstract: A replacement nozzle and heater sleeve for a nuclear reactor coolant system pressurizer. The heater, damaged heater sleeve exterior of the pressurizer and a portion of the heater sleeve inside the heater sleeve bore are removed. The heater sleeve bore is partially tapped to provide threads at its lower end. A sleeve having a first seal ring positioned at its upper end is threaded into the heater sleeve bore such that the first seal ring presses against the original heater sleeve and provides a seal. A second seal ring on a flange on the sleeve is pressed into sealing engagement with the exterior of the pressurizer.

Abstract: A robotic device for remotely inspecting pressurizer heater wells is provided which has the advantages of quickly, precisely, and reliably acquiring data at reasonable cost while also reducing radiation exposure of an operator. The device comprises a prober assembly including a probe which enters a heater well, gathers data regarding the condition of the heater well and transmits a signal carrying that data; a mounting device for mounting the probe assembly at the opening of the heater well so that the probe can enter the heater well; a first motor mounted on the mounting device for providing movement of the probe assembly in an axial direction; and a second motor mounted on the mounting device for providing rotation of the probe assembly. This arrangement enables full inspection of the heater well to be carried out.

Abstract: A method and apparatus for replacing a nozzle or heater sleeve in a nuclear reactor coolant system pressurizer. The tubular main body portion of the replacement nozzle is threadably received in the nozzle bore after the original heater and damaged heater sleeve have been removed and the nozzle bore partially tapped to provide a threaded area therein. The first or upper end of the main body portion is positioned substantially flush with the interior of the pressurizer. A seal weld is provided between the upper end of the main body portion and the interior of the pressurizer. A flange that extends radially from and adjacent the second or lower end of the main body portion bears against the exterior of the pressurizer when the main body portion is in its installed position inside the heater sleeve bore.

Abstract: The heating rod (3c) is cut from inside the casing (2) of the pressurizer in at least one zone (18, 19) by means of a cutting operation controlled remotely, and at least one portion (20a, 20b, 20c) of the rod (3c) is extracted by way of an inspection port (7) of the casing (2). The invention also relates to robotized cutting devices making it possible to carry out the cutting of heating rods (3) inside the casing (2) of the pressurizer (1).

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

Abstract: A replacement heater sleeve for a nuclear reactor coolant system pressurizer and method for replacing a damaged original heater nozzle. The original heater and heater nozzle are removed from the pressurizer and the original bore is enlarged. An outer sleeve is installed in the enlarged bore on the same center as the original heater nozzle and is substantially flush with the interior of the pressurizer. The outer sleeve is welded to the pressurizer on its interior and exterior surfaces. An inner sleeve is installed in the inner diameter of the outer sleeve and extends beyond the outer sleeve into the pressurizer. The inner sleeve is welded to the lower end of the outer sleeve and is provided with an inner diameter sized to receive a heater of the same size as that originally installed in the pressurizer. The inner sleeve also maintains the original heater alignment in the pressurizer.

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

Abstract: To ensure the sealed fixing of a replacement heating rod (4) on a collar (3) for passing through the casing (1) of a pressurizer, a weld is made in a single pass by a regulated MIG process. A shaft is fixed on one end of the heating rod projecting in relation to the collar (3) and in the extension thereof. A welding device is engaged and fixed in position on the shaft, this welding device comprising a tubular support (14) and an inert-gas orbital welding torch (10) with an electrode made of fusible metal, rotatably mounted about the axis of the support (14) which coincides with the axis of the rod (4). The welding torch (10) and a device to unwind welding wire which is placed in a cabinet (6) are set into synchronous rotation during the welding, by appropriate drive means. The welding is carried out in a single pass on the entire periphery of the heating rod (4) and of the collar (3).

Abstract: A water cooled nuclear reactor comprises a reactor core, a primary water coolant circuit and a pressurizer arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper chamber and a lower chamber by a casing, the reactor core and primary coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber.A plurality of pipes interconnect a steam space of the pressuriser with an upper portion of the primary coolant circuit via ports in the casing. A plurality of re-entrant surge ports interconnect a water space of the pressuriser with a lower portion of the primary coolant circuit. The surge ports have low flow resistance for water from the water space to the primary coolant circuit and high flow resistance in the opposite direction.

Abstract: A water-cooled direct cycle nuclear plant including a nuclear reactor, a turbine, a condenser, a purifying means and a feed water heater successively arranged as main constitution, further comprises means for measuring iron concentration in cooling water, and means for injection iron into cooling water for controlling iron amount in cooling water at an optimum level. The iron amount to be injected into cooling water is calculated based on the measured iron concentration so as to make the iron accumulation rate on fuel rod to be not less than 0.5 mg/m.sup.2 /hr.

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 accumulator (28) for use in a substantially zero gravity environment which utilizes the surface tension of a liquid to form a meniscus which acts as a gas-liquid interface providing a barrier to the passage or entrainment of a gas into the liquid. The apparatus includes a plurality of tubes (36) for capillary containment of a liquid, a grid member (34) provided with a plurality of holes (one of each tube) for receiving therein in sealing engagement an end of the tube. An opposite end of each of the tubes contains a gas. The apparatus further includes a housing (30) extending circumferentially about the grid member and defining a liquid zone (38) on the side of said grid member opposite said tubes located within the liquid zone and extending partially into each of the tubes is a body of liquid which forms a meniscus (42) in each of the tubes. Fluid communication is provided between the liquid zone of the apparatus and a source of liquid subject to thermal expansion and contraction.

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: Method for regulating the pressure of the primary circuit of a pressurized water nuclear reactor during shut-down phases, by use of an installation comprising, as a branch circuit to the primary circuit, a volume control circuit with a discharge valve and a charging valve. When the liquid level in the pressurizer approaches the top, a constant flow rate is maintained at the charging valve, the flow rate of the discharge valve is regulated by direct measurement of the primary pressure, successive sprinklings through the valve follow and the reduction in discharge flow rate is detected until further sprinkling has no further effect on the discharge flow rate.

Abstract: A pressure control system for a pressurized water nuclear reactor and method for quickly closing the valves associated with the pressurizer thereof has a temperature detection device on the loop seal of each valve and a valve responsive thereto to change water to the loop seal upon a temperature rise therein and stop the flow of water to the loop seal upon a temperature drop therein. The rapid formation of water seals in the loop seals for the valves protects the valve seats from wear and degradation and prolongs the life thereof.

Abstract: An improved pressure control system for a pressurized water nuclear reactor plant contains a novel, two stage sparger in the pressurizer relief tank. The two stage sparger has a primary conduit and secondary conduit, both of which have orifices through the walls thereof, and an interconnecting valve that is responsive to a pressure differential between the two conduits. The secondary conduit is preferably of a diameter less, but with a larger area of flow therefrom, than the primary conduit and is bifurcated, with two leg sections extending back towards the primary conduit.

Abstract: The overflow in the secondary of a steam generator of a nuclear-reactor plant, which occurs when one or more primary tubes conducting the coolant are ruptured, is controlled. The secondary of the steam generator, which contains the water that is converted to steam, is connected through valves to the pressurizer relief tank. The level of the liquid in the shell is monitored. An alarm is sounded when the level of the liquid in the generator reaches a predetermined height alerting the operator. When the level reaches a greater height, the valves are opened and the liquid in the shell is discharged into the pressurizer relief tank. The liquid in the tank is discharged into the containment sump when a disc is ruptured. The time taken after the rupture is 30 minutes as mandated by safety regulations. If, after the liquid in the shell reaches the level at which the valves are opened, the level drops to a height at which steam may be injected into the pressurizer relief tank, the valves are automatically closed.

Abstract: A reflective type insulation is utilized to elevate the temperature in the loop seal of a pressurizer. Damage to safety valve components and downstream piping caused by water in the piping is substantially reduced by exposing the loop seal piping to a portion of the outer surface of the pressurizer. Safety valve allowable sustained operating temperature requirements are considered in the insulation system design.

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.

Abstract: A model steam generator for simulating the environment inside a nuclear steam generator in order to monitor the condition of the heat exchange tubes of the nuclear steam generator is disclosed herein. The model steam generator includes an electrically-powered thermosyphon heating means for circulating a flow of vapor and condensate through the sample tube in a closed loop. The thermosyphon includes riser tubes which are concentrically disposed within the sample heat exchange tube along a substantial portion of the longitudinal axis thereof for both directing a stream of steam into the inside walls of the heat exchange tubes, and for insulating this flow of steam from the resulting stream of condensate which flows down from the inner walls of the sample tubes. The elimination of the counter-current forces between the stream of steam and the downflowing film of condensate prevents the formation of heat-obstructing ripples of condensate on the inner walls of the sample heat exchange tubes.

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: The invention relates to a secondary heat transfer circuit for a liquid metal-cooled nuclear reactor, as well as to a steam generator which can be used in such a circuit.According to the invention, a downstream buffer tank and an upstream buffer tank are arranged within the steam generator. The upstream buffer tank is annular and surrounds and communicates with an area of the generator by which the liquid metal flows downwards between the exchange zone and the discharge tube. The pressure of the neutral gas pocket in the downstream buffer tank is higher than that of the pocket of the upstream buffer tank.Application to fast neutron nuclear reactors.

Abstract: Apparatus for reducing the pressure in a pressurizer of a pressurized water type cooling system for a nuclear reactor when a predetermined pressure value is reached including a closed tank which is provided with coolant fluid outlet and inlet pipes or conduits which extend respectively from the bottom and top of the tank, and which are connectable, via a normally closed self-actuated mechanical pressure relief valve, to a spray nozzle in the upper end of the pressurizer and to the vapor space within the upper portion of the pressurizer respectively. The tank is disposed vertically above the pressurizer and is partially filled with liquid coolant which is subcooled for the pressure existing in the pressurizer, while the four-way valve is set to open in response to the predetermined pressure being reached in the pressurizer.

Abstract: The invention relates to a secondary heat transfer circuit for a liquid metal nuclear reactor.Each loop of the main circuit has in order a steam generator, a pump, and at least one heat exchanger positioned in the reactor vessel. A downstream buffer tank is located in the pipe connecting the generator to the pump, whilst the upstream buffer tank can be positioned either in the generator, or outside the latter.Application to the generation of electric power by means of a fast neutron reactor.

Abstract: The invention pertains to a containment system for a nuclear power plant which provides protection in the event of a loss of coolant accident or other malfunctions wherein the stored energy within the components of the system is contained and absorbed by the use of a plurality of coolant reservoirs capable of absorbing steam and supplying coolant to the reactor coolant system and feedwater to the secondary systems. The system includes a plurality of steam generators which may selectively be employed as an energy source during cooldown, and the design of the reservoirs provides high initial energy absorption as well as extended energy dissipation during cooldown.