A depressurization and cooling system for steam and/or condensable gases located in a containment. The system contains a steam condenser having an upstream port connected to the containment through an exhaust line and a downstream port connected to the containment through a backfeed line. The backfeed line contains a backfeed compressor. A re-cooling system for re-cooling the steam condenser is provided. The depressurization and cooling system is effective for re-cooling of the steam condenser. Accordingly, this is achieved as the re-cooling system is self-sustainable.

A Stirling engine provides a means to use the thermal energy in the sealed containment environment of a nuclear reactor building to provide emergency cooling. Acting as the prime mover in a coupled heat exchanger system, a Stirling engine could develop fluid flow thereby resulting in forced convection vice natural circulation and would not rely on an external power source during an unusual accident event where no electric power is available.

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.

The invention concerns a liquid metal-cooled fast-neutron nuclear reactor (1), comprising a system (2) for removing at least part of both the nominal power and the residual power of the reactor, which ensures, at the same time: removal of the residual power in a totally passive manner from the initial instant of the accident; removal of the heat through the primary vessel; implementation of a final cold source (container with PCM) other than the sodium/air or NaK/air heat exchangers used in the prior art.

To include a cylindrical vortex chamber 35, a small flow-rate pipe 37 connected to a peripheral plate 35C of the vortex chamber 35 along a tangential direction thereof, a large flow-rate pipe 36 connected to the peripheral plate 35C with a predetermined angle with respect to the small flow-rate pipe 37, an outlet pipe connected to an outlet 39 formed in a central part of the vortex chamber 35, and a straightening plate 50 that is arranged in a part between the outlet 39 and the peripheral plate 35C of the vortex chamber 35, and when jets flow into the vortex chamber 35 from the small flow-rate pipe 37 and the large flow-rate pipe 36, straightens impinging jets from the small flow-rate pipe 37 and from the large flow-rate pipe 36 having flowed into the vortex chamber 35 toward the outlet 39.

A high reliable water injection device is provided that injects water into a reactor containment vessel and can reliably shut off cooling water at normal times and quickly and reliably inject water into the reactor containment vessel without the need for external power, in a case of emergency. The water injection device injects water into a reactor containment vessel includes a flow path through which cooling water is supplied; a disk that closes the flow path; a swing arm that is connected to the disk and performs closing and opening of the flow path by the disk; and a weight that is connected to the swing arm via a swing lever, in which the weight is supported by a support member made of a low melting point alloy.

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.

A safety injection device includes a containment building, a reactor coolant system provided inside the containment building and accommodating a reactor coolant formed to carry heat energy generated according to fission of fuel, a cooling water storage section accommodating cooling water injected into the reactor coolant system, a power producing section producing power with steam discharged from the reactor coolant system in case of an accident, a steam supply pipe transmitting steam discharged from the reactor coolant system to the power producing section, a steam discharge pipe discharging steam drove the power producing section and a safety injection line supplying cooling water accommodated in the cooling water storage section to the inside of the reactor coolant system. In addition, cooling water accommodated in the cooling water storage section is supplied to the inside of the reactor coolant system, based on the power produced by the power producing section, through a cooling water inlet pipe connecting the cooling water storage section and the power producing section, the cooling water discharge pipe connecting the power producing section and the safety injection line, and the safety injection line.

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.

A nuclear reactor is configured with an intermediate coolant loop for transferring thermal energy from the reactor core for a useful purpose. The intermediate coolant loop includes a bypass flowpath with an air heat exchanger for dumping reactor heat during startup and/or shutdown. A fluidic diode along the bypass flowpath asymmetrically restricts flow across the bypass flowpath, inhibiting flow in a first flow direction during a full power operating condition and allowing a relatively uninhibited flow in a second direction during a startup and/or shut down low power operating condition.