Abstract: A cooling system in a nuclear power plant is disclosed, including a boundary section disposed inside a containment to enclose a reactor coolant system, and configured to restrict steam containing radioactive materials generated in the reactor coolant system from leaking into paths other than a discharge part. An In-Containment Water Storage Tank (IRWST) is disposed outside the boundary section and is configured to store refueling water therein. An emergency cooling tank is disposed outside the containment and is provided with a condensation heat exchanger. A gas-liquid separator is connected to the emergency cooling tank outside the containment. A return line is configured to connect the gas-liquid separator and the boundary section such that condensate generated by condensing the steam within the boundary section, through the emergency cooling tank and the gas-liquid separator, is discharged toward the boundary section upon an occurrence of a nuclear power plant accident.

Abstract: Disclosed herein is a nuclear power plant main steam system that reduces the atmospheric discharge of radioactive materials generated in an accident. The system includes: a decontamination water tank containing decontamination water; and a connection pipe for connecting the decontamination water tank to a main steam pipe which connects a steam generator and a turbine. A main steam safety valve or a connection valve is provided as a three-way valve configured to discharge the generated steam to the atmosphere when an accident occurs within a design basis and to transfer the generated steam to the decontamination water tank when an accident involving damage to nuclear fuel occurs. The main steam system reduces discharge of radioactive materials to the atmosphere when a containment bypass accident (e.g., a steam generator tube rupture caused by high-temperature steam) occurs.

Abstract: The present invention relates to a prevention device for loss of coolant accident (LOCA) and a nuclear reactor having the same. The prevention device for LOCA includes a nozzle portion integrally formed in a reactor vessel and having a communication hole communicating with the inside of the reactor vessel, a nozzle finishing portion assembled to the nozzle portion and an injection line for injecting a fluid to the inside of the reactor vessel respectively on both sides thereof in a communicating manner, and a check valve mounting portion installed to be embedded inside the nozzle portion and having at least one check valve opened by flow such that the fluid is injected into the reactor vessel, wherein the check valve blocks outflow of a reactor coolant from the reactor vessel in case of failure of the injection line.

Abstract: A nuclear plant includes a nuclear reactor, a containment structure that at least partially defines a containment environment of the nuclear reactor, and a passive containment cooling system that causes coolant fluid to flow downwards from a coolant reservoir to a bottom of a coolant channel coupled to the containment structure and rise through the coolant channel toward the coolant reservoir due to absorbing heat from the nuclear reactor. A check valve assembly, in fluid communication with the coolant reservoir, selectively enables one-way flow of a containment fluid from the containment environment to the coolant reservoir, based on a pressure at an inlet being equal to or greater than a threshold magnitude. A fusible plug, in fluid communication with the coolant reservoir at a bottom vertical depth below the bottom of the coolant reservoir, enables coolant fluid to flow into the containment structure based on at least partially melting.

Abstract: An integrated passive cooling containment structure for a nuclear reactor includes a concentric arrangement of an inner steel cylindrical shell and an outer steel cylindrical shell that define both a lateral boundary of a containment environment of the nuclear reactor that is configured to accommodate a nuclear reactor and an annular gap space between the inner and outer steel cylindrical shells, a concrete donut structure at a bottom of the annular gap space, and a plurality of concrete columns spaced apart azimuthally around a circumference of the annular gap and extending in parallel from a top surface of the concrete donut structure to a top of the annular gap space. The outer and inner steel cylindrical shells and the concrete donut structure at least partially define one or more coolant channels extending through the annular gap space.

Abstract: A method and apparatus of limiting power of a boiling water nuclear reactor system includes a reactor pressure vessel, a reactor core disposed in the reactor pressure vessel, a core shroud surrounding the reactor core, a downcomer region disposed between an inner surface of the reactor pressure vessel and the core shroud, a steam line connected to an upper end of the reactor pressure vessel and a condenser system that receives steam from the reactor pressure vessel. A portion of the condenser system condensate is returned to the reactor pressure vessel of the boiling water reactor inside the core barrel above the core rather than into the downcomer. Returning the condensate in this way increases the effectiveness of an isolation condenser system or if the condensate is a portion of the feedwater from the main condenser it provides an effective means to regulate core flow and core power.

Abstract: A flange flush tool includes an outer tube having a first end and a second end, the first end configured to engage the flange, and a shield tube within the outer tube. A channel is defined between an outer surface of the shield tube and an inner surface of the outer tube. The shield tube is configured to move longitudinally within the outer tube.

Abstract: According to an embodiment, an emergency core cooling system has: three active safety divisions each including only one motor-driven active safety system; one passive safety division including a passive safety system; an emergency power source disposed in each of the active safety divisions to supply electric power to the motor-driven active safety system; and an advanced passive containment cooling system disposed in the passive safety division. Only two active safety divisions each includes a low pressure flooder system that is commonly used with a residual heat removal system as the only one motor-driven active safety system. The other active safety division includes an air-cooled injection system as the only one motor-driven active safety system.

Abstract: A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel having a lower portion containing a nuclear reactor core and a vessel head defining an integral pressurizer. A reactor coolant pump (RCP) mounted on the vessel head includes an impeller inside the pressure vessel, a pump motor outside the pressure vessel, and a vertical drive shaft connecting the motor and impeller. The drive shaft does not pass through the integral pressurizer. The drive shaft passes through a vessel penetration of the pressure vessel that is at least large enough for the impeller to pass through.

Abstract: Nuclear reactors include isolation condenser systems that can be selectively connected with the reactor to provide desired cooling and pressure relief. Isolation condensers are immersed in a separate chamber holding coolant to which the condenser can transfer heat from the nuclear reactor. The chamber may selectively connect to an adjacent coolant reservoir for multiple isolation condensers. A check valve may permit coolant to flow only from the reservoir to the isolation condenser. A passive switch can operate the check valve and other isolating components. Isolation condensers can be activated by opening an inlet and outlet to/from the reactor for coolant flow. Fluidic controls and/or a pressure pulse transmitter may monitor reactor conditions and selectively activate individual isolation condensers by opening such flows. Isolation condenser systems may be positioned outside of containment in an underground silo with the containment, which may not have any other coolant source.

Abstract: The application provides a self-diagnosis and accident-handling unmanned nuclear reactor, which: can passively cool down excessively generated heat without an operation of an operator when a malfunction of the nuclear reactor has occurred, wherein a cooling operation for safety measures can be carried out in a completely passive manner without a separate control command by a change in environmental conditions such as the structure and pressure of the nuclear reactor; and has a simpler structure compared to that of a conventional nuclear reactor safety system. It also provides a self-diagnosis and accident-handling unmanned nuclear reactor, which performs heat exchange by using a two-phase heat transfer mechanism, wherein heat exchange performance is maximized by introducing a spray-type heat exchanger having an optimized structure in which channels are three-dimensionally arranged, and can also easily and passively control heat exchange without a separate control means by using saturated steam pressure.

Abstract: When a power source is lost after an operation stop of a nuclear power plant, a first open/close valve is opened via a first battery at an early stage and steam in a reactor pressure vessel (RPV) is condensed in a suppression pool. The heat of the water in the suppression pool is transmitted to a cooling water pool located below inner space between first and second reactor containment vessels surrounding the RPV. A second open/close valve is opened via a second battery at the early stage and cooling water in a tank is injected into the RPV. After the early stage, a third open/close valve is opened via a third battery, and a cooling medium becomes steam by an evaporator in the RPV, the steam being condensed by a condenser disposed in the inner space to become a liquid of the cooling medium and is returned to the evaporator.

Abstract: In a pressurized water reactor (PWR), emergency core cooling (ECC) responds to depressurization due to a vessel penetration break at the top of the pressure vessel by draining water from a body of water through an injection line into the pressure vessel. A barrier operates concurrently with the ECC to suppress flow of liquid water from the pressure vessel out the vessel penetration break. The barrier may comprise one or more of: (1) an injection line extension passing through the central riser to drain water into the central riser; (2) openings in a lower portion of a central riser to shunt some upward flow from the central riser into a lower portion of the downcomer annulus; and (3) a surge line providing fluid communication between a pressurizer volume at the top of the pressure vessel and the remainder of the pressure vessel which directs water outboard toward the downcomer annulus.

Abstract: A nuclear reactor chamber comprises an inlet portion. The chamber is a part of a nuclear power plant. At least one container contains liquid nitrogen and cold nitrogen vapor and includes an outlet portion. At least one thermally activated release mechanism is respectively connected between one of the at least one container and the inlet portion. Each thermally activated release mechanism is configured to release the liquid nitrogen from a connected container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber.

Abstract: An atomic power plant is provided including a primary containment vessel (PCV) 1, a reactor pressure vessel (RPV) 3, a main steam line 4, two steam safety relief valves (SRVs) 6, a pressure suppression pool (S/P) 8, an SRV exhaust pipe 9 which is connected to a quencher 10, a temperature measuring instrument 12 which measures a temperature inside the quencher 10, an SRV controller 13 which controls opening and closing of the SRVs 6. After a lapse of predetermined time from when the SRV 6 is opened, in a case where it is determined that a temperature detected by the temperature measuring instrument 12 is equal to or smaller than a predetermined threshold value, the SRV controller 13 causes the SRV 6 to which the temperature measuring instrument 12 detecting the temperature leads, to be closed and to be prohibited from being opened.

Abstract: The present invention relates to a passive heat removal system which circulates cooling fluid to a steam generator via a main water supply line connected to the lower inlet of the steam generator, and a main steam pipe connected to the top outlet of the steam generator, to remove sensible heat of a nuclear reactor coolant system and residual heat of a core. The heat removal system comprises supplementary equipment for receiving surplus cooling fluid or for supplying supplementary cooling fluid in order to maintain the flow rate of the cooling fluid within a predetermined range. The supplementary equipment comprises: a supplementary tank, installed at a height between the lower inlet and the top outlet of the steam generator; a first connection pipe, connected to the main steam pipe and the supplementary tank; and a second connection pipe, connected to the supplementary tank and the main water supply pipe.

Abstract: A passive containment cooling and filtered venting system includes: an outer well; a scrubbing pool arranged in the outer well; a cooling water pool installed above the dry well and the outer well; a heat exchanger partly submerged in the cooling water; a gas supply pipe that is connected to the inlet plenum of the ruin of the heat exchanger at one end and connected to a gas phase region of the containment vessel at the other end; a condensate return pipe that is connected to the outlet plenum of the heat exchanger at one end, and connected to inside the containment vessel at other end; and a gas vent pipe that is connected to the outlet plenum of the heat exchanger at one end and is submerged in the scrubbing pool at other end.

Abstract: A pressure-relief system for the containment of a nuclear power facility allows reliable operation of a wet scrubber for the pressure relief flow with a simultaneously compact structural design. The pressure relief system has a pressure relief line guided through the containment and can be closed by a shut-off valve, a wet scrubber arranged in a portion of the pressure relief line located inside the containment, for the pressure relief flow which forms in the pressure-relief mode when the shut-off valve is open, a reservoir arranged inside the containment and is fluidically connected to the remaining inner space of the containment such that any overpressure, with respect to the surroundings outside the containment, prevailing in the containment is transferred at least in part to the reservoir, and a supply line leading from the reservoir to the wet scrubber for supplying the wet scrubber with fluid from the reservoir.

Abstract: The present invention provides a facility for reducing radioactive material comprising: a cooling water storage unit installed inside a containment and formed to store cooling water; a boundary unit forming a boundary of radioactive material inside the containment and surrounding a reactor coolant system installed inside the containment to prevent a radioactive material from releasing from the reactor coolant system or a pipe connected with the reactor coolant system to the containment; a connecting pipe connected with an inner space of the boundary unit and the cooling water storage unit to guide a flow of a fluid caused by a pressure difference between the boundary unit and the cooling water storage unit from the boundary unit to the cooling water storage unit; and a sparging unit disposed to be submerged in the cooling water stored in the cooling water storage unit and connected with the connecting pipe to sparge the fluid that has passed through the connecting pipe and the radioactive material contained i

Abstract: A nuclear power plant according to an embodiment comprises: a reactor well; a reactor well upper lid; an operation floor; an operation floor area wall; a standby gas treatment system; and a reactor well exhaust section to release the gas inside the reactor well to the environment without releasing the gas into the operation floor area in an event of a predetermined accident, e.g., causing diminished cooling of a containment vessel or otherwise increasing its temperature. The standby gas treatment system includes: a suction pipe to take in gas inside the reactor building; an exhaust fan; a standby gas treatment system exhaust pipe; a heater that is disposed between the suction pipe and the standby gas treatment system exhaust pipe; and a filter to filter the gas heated by the heater and to send the gas to the standby gas treatment system exhaust pipe.

Abstract: In accordance with the present invention, there is provided a strainer system for use in a nuclear sump. The strainer system of the present invention includes at least one primary strainer module which defines a primary strainer/filter surface. In the strainer system, the primary strainer surface of the primary strainer module has a debris interceptor which is cooperatively engaged thereto, and may be outfitted with one or more pressure released or activated membranes. In a loss of coolant accident, the debris interceptor, alone or in combination with the pressure activated membrane(s), is adapted to reduce the differential pressure experienced across the strainer system in nuclear power plants with medium to high fiber loads.

Abstract: A cleansing system for improving operation of a petrochemical plant or refinery. The petrochemical plant or refinery may include a fractionation column, a condenser, and a pump. Equipment, such as condensers, receivers, reboilers, feed exchangerss, and pumps may be divided into subsections. Temperatures, pressures, flows, and other plant operations may be used for optimizing plant performance. A cleansing unit performs an enhanced cleansing process, which may allow early detection and diagnosis of the plant operating conditions based on one or more environmental factors.

Abstract: A component cooling water system for a nuclear power plant. In one embodiment, the system includes an inner containment vessel housing a nuclear reactor and an outer containment enclosure structure. An annular water reservoir is formed between the containment vessel and containment enclosure structure which provides a heat sink for dissipating thermal energy. A shell-less heat exchanger is provided having an exposed tube bundle immersed in water held within the annular water reservoir. Component cooling water from the plant flows through the tube bundle and is cooled by transferring heat to the annular water reservoir. In one non-limiting embodiment, the tube bundle may be U-shaped.

Abstract: According to an embodiment, a nuclear power plant has a core; a reactor pressure vessel; a dry well; a wet well; a vacuum breaker; a containment vessel including the dry well, the LOCA vent pipe, the wet well, and the vacuum breaker; a cooling water pool placed outside the containment vessel; a heat exchanger at least partially submerged in cooling water; a gas supply pipe connected to the inlet plenum of the heat exchanger and the dry well; a condensate return pipe connected to the outlet plenum of the heat exchanger and the containment vessel; and a gas vent pipe connected to the outlet plenum of the heat exchanger and an outside of the wet well so that non-condensable gas inside the heat exchanger is released out of the wet well. The gas vent pipe is not connected to the wet well.

Abstract: A containment for a water cooled and moderated nuclear reactor incorporates two or more separate containment zones. These zones are constructed in such a manner that a leak or break in the reactor coolant system located within one zone will remain confined within this particular zone, so that no adverse ambient conditions of pressure, temperature, and humidity will propagate to any of the other zones. The separation between zones is achieved by having a partition plate extending between the containment envelope and the reactor coolant system where the partition plate is attached to one of the main components of the reactor coolant system. For example, this can be the reactor pressure vessel, as shown in some of the embodiments. The partition is designed to the same pressure and temperature conditions as the containment vessel envelope to ensure a leak tight and permanent separation between adjacent zones.

Abstract: A safety system for a nuclear plant includes a plurality of catalytic recombiner elements each triggering a recombination reaction with oxygen when hydrogen is entrained in an onflowing gas flow, so that reliable elimination of the hydrogen from the gas mixture is ensured with an especially high degree of operational safety even based on comparatively extreme conditions or scenarios of the aforementioned type. The recombiner elements and/or the flow paths each connecting two recombiner elements on the gas side are configured in such a way that a pressure pulse triggered in the gas medium by an ignition during the recombination reaction in a first recombiner element triggers a gas displacement process having a flow rate of at least 5 m/s in the onflow region of a second, adjacent recombiner element. A nuclear plant with a safety system is also provided.

Abstract: A condensation chamber cooling system of a condensation chamber for a boiling water reactor has a heat exchanger outside the condensation chamber. An elongate cooling module is provided in the condensation chamber with an evaporation space in its upper region. The cooling module is configured such that the evaporation space is located above a maximum filling level of the condensation chamber. The cooling module includes at least one riser pipe and one downpipe that issue with their upper ends into the evaporation space and with their lower ends in the condensation chamber. A first pressure line leads from the evaporation space to the heat exchanger and, from there, a second pressure line which issues in the condensation chamber below the minimum filling level. Thus, the condensation chamber, the pressure lines, the cooling module and the heat exchanger form a passive closed cooling circuit.

Abstract: In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer.

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

Abstract: The present invention relates to an alternate feedwater injection system to at least partially mitigate the effects of an aircraft impact on a light water nuclear reactor positioned in a reactor building. The light water nuclear reactor has a primary system and a reactor core. The alternate feedwater injection system includes a water storage tank, an injection point into the primary system, a pump capable to transfer water from the water storage tank to the injection point and ultimately to the reactor core. The water storage tank and pump are located external to a reactor building and outside of an identified aircraft impact area or inside the identified aircraft impact area and provided with a means of protection from the aircraft impact.

Abstract: The present invention is directed to remote operation of an operation valve such as an air operated valve even at the time of power loss. A gas supply apparatus of the present invention includes: an operation valve mounted in some midpoint of a piping for passing at least gas in a plant and operating a valve body by the gas flowing in the piping; an solenoid valve mounted in some midpoint of the piping and allowing/stopping flow of the gas to the operation valve; and a gas supply source for supplying gas to the solenoid valve. A switching valve for switching between exhaust from the solenoid valve and gas supply to the solenoid valve is mounted in an exhaust line of the solenoid valve and, at the time of power loss, the switching valve is switched to connection to the gas supply source for supplying gas to the solenoid valve.

Abstract: A passive cooling system of a pressurized water reactor that relies on a depressurization system to reduce the pressure in the reactor vessel in the event of a loss of coolant accident and vent the steam generated by the decay heat of the reactor core in a post loss of coolant accident stage. The depressurization results in a low pressure difference between the reactor vessel and the containment and enables gravity driven cooling system injection into the reactor vessel. The depressurization system includes a flow restrictor within an orifice in the reactor vessel wall that connects to a vent pipe which forms a flow path between the interior of the reactor vessel and the containment atmosphere when a valve within the vent pipe is in an open position. Preferably, the flow restrictor is a venturi that has a gradual contraction and a gradual expansion in the flow path area.

Abstract: A power module assembly includes a reactor core immersed in a coolant and a reactor vessel housing the coolant and the reactor core. An internal dry containment vessel submerged in liquid substantially surrounds the reactor vessel in a gaseous environment. During an over-pressurization event the reactor vessel is configured to release the coolant into the containment vessel and remove a decay heat of the reactor core through condensation of the coolant on an inner surface of the containment vessel.

Abstract: According to an embodiment, a reactor pressure vessel depressurization system has: a main steam safety relief valve, a main steam safety relief valve driving gas pipe; a three-way solenoid valve having the first connection port; a driving gas feed pipe connected to the second connection port; and a containment vessel external connection pipe connected to the third connection port and extending to outside of the reactor containment vessel. The three-way solenoid valve is either in the first communication state where the first connection port communicates with the second connection port or in the second communication state where the first connection port communicates with the third communication port. The containment vessel external connection pipe has an open communication section open in normal operation and capable of being unopened, and an external gas receiving section capable of receiving second driving gas.

Abstract: A method and a device depressurize a nuclear power plant. A depressurization flow is conducted out of a containment shell into the atmosphere via a depressurization line having a filter system. The filter system contains a filter chamber having an inlet, an outlet, and a sorbent filter. The depressurization flow is first conducted in a high-pressure section, then is depressurized by expansion at a throttle device, then conducted through the filter chamber having the sorbent filter, and finally blown out. To enable an effective retention of activity carriers contained in the depressurization flow, including organic compounds containing iodine, the depressurization flow depressurized by the throttle device is conducted through a superheating section before the depressurization flow enters the filter chamber, in which superheating section the depressurization flow is heated from the not yet depressurized depressurization flow to a temperature that is at least 10° C. above the dew point temperature.

Abstract: A system and a method for a passive containment vent system for a Boiling Water Reactor (BWR). The system is capable of venting and scrubbing a gaseous discharge from the primary containment of the BWR over a prolonged period of time leading up to or following a serious plant accident, without the need for monitoring by on-site plant personnel. External electrical power is not required (following initial activation of the system) in order to operate the containment vent system. The system may protect the integrity of primary containment during and following the serious plant accident.

Abstract: The present invention allows to inhibit antiplane deformation of steel plates due to the back pressure generated by water content of the concrete, and allows observation of the concrete during a service period. In a steel concrete structure including an inner steel plate exposed to a high temperature environment, an outer steel plate, studs and concrete, a steam discharging pipe and a pipe opening are provided. The steam discharging pipe includes small openings. Water content inside the concrete evaporates in a state the temperature of the concrete reaches approximately 100° C., becomes steam, is concentrated in the vicinity of the rear surface of the steel plate, and is discharged to the outside of a containment vessel from the small openings through the steam discharging pipe.

Abstract: A power module assembly includes a reactor vessel containing a reactor core surrounded by a primary coolant. A containment vessel is adapted to be submerged in a containment cooling pool and to prohibit a release of the primary coolant outside of the containment vessel. A secondary cooling system is configured to remove heat generated by the reactor core. The heat is removed by circulating liquid from the containment cooling pool through the primary coolant.

Abstract: A thermal sleeve is mechanically attached to the bore of a surge nozzle of a pressurizer for the primary circuit of a pressurized water reactor steam generating system. The thermal sleeve is attached with a series of keys and slots which maintain the thermal sleeve centered in the nozzle while permitting thermal growth and restricting flow between the sleeve and the interior wall of the nozzle.

Abstract: A nuclear plant has a containment shell and a pressure relief pipe connected thereto in which a blowing device and a Venturi washer placed in a container with a washing liquid are connected in series. Even the finest particles or aerosols carried by air are held in the Venturi washer with a very high degree of reliability and the release thereof in environment is excluded in a particularly reliable manner in the case of decompression even associated with seal failures. For this purpose, the size of the blowing device and the Venturi washer are selected in such a way that during the operation of the blowing device a flow rate of liquid in the Venturi washer flowing to the decompressing pipe is higher than 130 m/sec, preferably higher than 180 m/sec.

Abstract: A nuclear plant has a containment shell and a pressure relief pipe connected thereto in which a blowing device and a Venturi washer placed in a container with a washing liquid are connected in series. Even the finest particles or aerosols carried by air are held in the Venturi washer with a very high degree of reliability and the release thereof in environment is excluded in a particularly reliable manner in the case of decompression even associated with seal failures. For this purpose, the size of the blowing device and the Venturi washer are selected in such a way that during the operation of the blowing device a flow rate of liquid in the Venturi washer flowing to the decompressing pipe is higher than 130 m/sec, preferably higher than 180 m/sec.

Abstract: According to an embodiment, a nuclear power plant has a core; a reactor pressure vessel; a dry well; a wet well; a vacuum breaker; a containment vessel including the dry well, the LOCA vent pipe, the wet well, and the vacuum breaker; a cooling water pool placed outside the containment vessel; a heat exchanger at least partially submerged in cooling water; a gas supply pipe connected to the inlet plenum of the heat exchanger and the dry well; a condensate return pipe connected to the outlet plenum of the heat exchanger and the containment vessel; and a gas vent pipe connected to the outlet plenum of the heat exchanger and an outside of the wet well so that non-condensable gas inside the heat exchanger is released out of the wet well. The gas vent pipe is not connected to the wet well.

Abstract: A system includes a containment vessel configured to prohibit a release of a coolant, and a reactor vessel mounted inside the containment vessel. An outer surface of the reactor vessel is exposed to below atmospheric pressure, wherein substantially all gases are evacuated from within the containment vessel.

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: In one embodiment, the heat removal system includes a storage tank configured to store a heat transfer medium, a transfer system configured to selectively transfer the heat transfer medium from the storage tank to the nuclear reactor, and a delivery system operationally connected to the transfer system. The delivery system is configured to deliver the heat transfer medium to a suppression pool room of the nuclear reactor. The suppression pool room houses a suppression pool.

Abstract: A blocking device for preventing the actuation of an automatic depressurization system in a pressurized nuclear reactor system due to spurious signals resulting from a software failure. The blocking signal is removed when the coolant level within the core makeup tanks drop below a predetermined level.

Abstract: A passive cooling and depressurization system for a pressurized water nuclear plant is provided with a cooling water pool, a steam supply piping, a heat exchanger, a steam supply valve, a coolant return pipe and an outlet valve. The steam supply piping extends from the gas phase of the pressurizer. The heat exchanger exchanges heat between water stored in the cooling water pool and steam flowing through the steam supply piping. The steam supply valve is equipped on the steam supply piping. The coolant return pipe extends from the heat exchanger to a liquid phase of the reactor pressure boundary. The outlet valve is equipped on the coolant return pipe.

Abstract: A passive cooling and depressurization system for a pressurized water nuclear plant is provided with a cooling water pool, a steam supply piping, a heat exchanger, a steam supply valve, a coolant return pipe and an outlet valve. The steam supply piping extends from the gas phase of the pressurizer. The heat exchanger exchanges heat between water stored in the cooling water pool and steam flowing through the steam supply piping. The steam supply valve is equipped on the steam supply piping. The coolant return pipe extends from the heat exchanger to a liquid phase of the reactor pressure boundary. The outlet valve is equipped on the coolant return pipe.

Abstract: A safety valve drive system is operated such that a safety valve of a nuclear power plant is opened by supplying a driving gas by using a pilot valve at an occurrence of an accident or a transient state to thereby protect a reactor against pressure application. The safety valve drive system is provided with a safety valve drive unit, as a function of actuating the safety valve, and cables. The safety valve drive unit actuates in a manner that the safety valve is opened in response to respective auto-depressurization system actuating signals for two or more segments among respective auto-depressurization system actuating signals for four segments, and is closed if an auto-depressurization system actuating signal for one or less segment among the auto-depressurization system actuating signals for the four segments is received. The cables are connected to the safety valve drive unit and used to transfer the auto-depressurization system actuating signals for the four segments.

Abstract: A method and a device depressurize a nuclear power plant. A depressurization flow is conducted out of a containment shell into the atmosphere via a depressurization line having a filter system. The filter system contains a filter chamber having an inlet, an outlet, and a sorbent filter. The depressurization flow is first conducted in a high-pressure section, then is depressurized by expansion at a throttle device, then conducted through the filter chamber having the sorbent filter, and finally blown out. To enable an effective retention of activity carriers contained in the depressurization flow, including organic compounds containing iodine, the depressurization flow depressurized by the throttle device is conducted through a superheating section before the depressurization flow enters the filter chamber, in which superheating section the depressurization flow is heated from the not yet depressurized depressurization flow to a temperature that is at least 10 ° C. above the dew point temperature.