Abstract: A reactor containment facility having a reactor pressure vessel containing a core; a dry well in which the reactor pressure vessel is arranged; a suppression chamber holding suppression-pool water and forming above the suppression-pool water a wet well; and a plurality of vent pipes allowing the dry well to communicate with the suppression-pool water; a steel wall which is in contact with the suppression-pool water of the suppression chamber and which surrounds at least the pool water so as to form a containment vessel which houses the dry well and the suppression chamber; and an outer peripheral pool containing cooling water in contact with the outer peripheral surface of the steel wall.

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: A pressure control chamber of a reactor containment vessel for receiving the atmosphere of the reactor containment vessel containing leakage steam developing in the event of an accident includes a pressure control pool, and a wet well disposed above the pressure control pool. The wet well is divided into an inner peripheral portion and an outer peripheral portion. The inner and outer peripheral portions are communicated with each other via the pressure control pool. An outlet port for flowing the atmosphere of the reactor containment vessel containing the leakage steam is provided at the inner peripheral side of the pressure control pool. When the atmosphere of the reactor containment vessel containing the leakage steam is introduced into the pressure control pool in the event of the accident, this atmosphere is condensed by the water in the pressure control pool, and the gas of this atmosphere not condensed is accumulated in the inner peripheral portion of the wet well.

Abstract: A reactor containment vessel comprises typically a reactor pressure vessel housed in a dry well of the reactor containment vessel, a vent passage through which steam in the dry well is introduced into coolant accommodated in a pressure suppression chamber of the reactor containment vessel, a closed space formed at a position lower than the level of the normal liquid surface of the coolant, a first passage having an inlet opened into the pressure suppression chamber at a level higher than that of the normal liquid surface of the coolant and an outlet opened into the closed space, and a second passage communicating between the closed space and the dry well through a counter flow preventing arrangement for checking the flow directed toward the closed space, and therefore, noncondensable gas and liquid accumulated in the pressure suppression chamber are discharged into the dry well the pressure in which is higher than that in the pressure suppression chamber by making use of a difference in water head increased d

Abstract: A primary containment vessel, comprising a reactor pressure vessel in which a core is accommodated, a dry-well volume covering the upper portion of the reactor pressure vessel, a suppression chamber provided with a suppression pool surrounding an intermediate portion of the reactor pressure vessel, has a plurality of vent tubes establishing a connection between the dry-well volume and the suppression pool and an annular baffle plate disposed in the suppression pool along the side wall of the suppression chamber, having its top end positioned at a height between a free surface of the suppression pool and an outlet of the vent tube opened in the suppression pool and its lower end positioned at a height between the outlet and the bottom of the suppression pool. If a loss of coolant accident occurs, hot steam discharged into the dry-well volume is discharged into the suppression pool through the vent tube so as to heat water on the inside of the baffle plate.

Abstract: An emergency core cooling system for spraying cooling water onto a nuclear reactor core of a nuclear reactor includes at least one header disposed in the nuclear reactor and arranged at an upper and outer circumferential position with respect to the reactor core. A plurality of spray nozzles are mounted on the at least one header for spraying cooling water onto the reactor core. The spray nozzles have a center axis and at least one of the plurality of spray nozzles is mounted on the at least one header so that the center axis thereof extends upwardly with respect to a horizontal axis whereas at least one of the remaining plurality of spray nozzles has the center axis thereof extending downwardly with respect to the horizontal axis. Cooling water is conducted to the at least one header for enabling spraying of the water in the emergency situation.

Abstract: An emergency reactor core cooling structure comprises a flow passage provided on an upper core plate so as to extend downward therefrom into the core. The flow passage has an orifice member at a lower portion thereof. The orifice member has such a portion that the cross-sectional area of the flow passage is made minimal at the orifice member. The flow passage causes the cooling water held on the upper plate by vapor flowing up on an accident to flow down therethrough, whereby the core is cooled rapidly.

Abstract: A nuclear reactor start-up method of starting up a boiling water reactor which is in the cold shut-down state. The method includessetting a subcooled temperature of a coolant in a reactor core inlet to a range smaller than a critical subcooled temperature which is determined by the condition whereby the coolant starts boiling in the reactor core and the condition whereby the instability due to a periodic variation of buoyancy begins,thereafter starting an increase of power of the reactor.