Hitachi, Ltd.

Hitachi Engineering Co., Ltd.

Hitachi-ge Nuclear Energy, Ltd.

Michio Murase

Yoshiyuki Kataoka

Masanori Naitoh

Masataka Hidaka

Hiroaki Suzuki

Shigeo Hatamiya

Kenji Tominaga

Isao Sumida

Shozo Yamanari

Hidefumi Araki

Masao Chaki

Koji Hashimoto

Kotaro Inoue

Tetsuo Horiuchi

Hisamichi Inoue

Ryuhei Kawabe

Tsuyoshi Niino

Naoyuki Ishida

Toshitsugu Nakao

Masaki Matsumoto

Terufumi Kawasaki

Toshihiko Sugisaki

Shigeto Murata

Tohru Fukui

Shoichiro Kinoshita

Satoshi Miura

Toru Fukui

Akihiko Minato

Shouichiro Kinoshita

Yasutaka Iwata

Masanori Naioh

A steam separator comprises an outer main swirler and an inner auxiliary swirler which is smaller than the main swirler. The swirlers are provided so as to be concentric on the inner wall at the lower side of the first stage inner cylinder. In the steam separator, when the gas-liquid two-phase flow which flows in the vicinity of the axial center of the first stage inner cylinder passes the auxiliary swirler, it is separated into steam and water by the centrifugal force. The separated water (droplets) is introduced into the main swirler. When the separated water (droplets) passes the main swirler, it is separated at the inner wall side of the first stage inner cylinder by the centrifugal force. Pressure loss in a steam separator is reduced and steam separation capability is increased without increasing the moisture from the steam separator.

Steam separator, boiling water reactor and swirler assembly

Pipes are mounted horizontally on an outlet portion of a gas vent pipe in a condenser-type heat removal system, and have-openings which are formed only in a portion below a horizontal plane in which axes of these pipes lie, thereby enlarging a region where uncondensed steam is mixed with water in a suppression pool. With this arrangement, a surface portion of the water of the suppression pool is prevented from becoming hot, and the temperature of this pool water is uniformed. As a result, the temperature of the water surface of the suppression pool is lowered, thereby decreasing the pressure within a primary containment vessel. This enhances the reliability of the primary containment vessel, and reduces a design strength of the primary containment vessel.

Cooling system for primary containment vessel in nuclear power plant and

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. The facility further includes: a dividing structure for dividing the wet well of the suppression chamber into a first space which is in contact with the water surface of the suppression-pool water and a second space which is not in contact therewith; a first passage which allows the first space to communicate with the second space and which has an area smaller than that of the dividing device; and a cooling device for keeping the second space at a temperature lower than that of the first space.

Reactor containment facilities

In a boiling water nuclear reactor, a reactor primary cooling water line is filled with water at the start-up time, and the inside of a pressure vessel is pressurized by a pressurized tank and at the same time control rods are withdrawn to thereby heat cooling water in a state of single-phase flow to high temperature. Succeedingly, the pressurization is released and the pressure of the pressure vessel is gradually approximated to the saturation pressure corresponding to the cooling water temperature to thereby make the cooling water transit into a state of two-phase flow, and thereafter the cooling water is heated by nuclear reaction in the state of two-phase flow to thereby obtain predetermined reactor running temperature and pressure. This prevents the flow fluctuation and the reduction of stability of core due to occurrence of unstable phenomena of low temperature two-phase flow at the start-up time of the reactor to thereby make it possible to perform stable starting-up of the reactor and to shorten the start-up time of the reactor to thereby realize improvement in economy as well as reliability.

Boiling water nuclear reactor and start-up process thereof

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 due to increase of the liquid while allowing fluid to be condensed to newly flow into the pressure suppression chamber from the dry well so as to be condensed, over a long cooling period after the occurrence of an accident of piping fracture in the reactor containment vessel. Owing to condensation, the pressure in the reactor containment vessel can be reduced, and moreover, fluid can make a natural circulation in a closed circuit passing through the dry well, the pressure suppression chamber and the closed space in order without flowing outside the circuit.

Reactor containment vessel

A natural circulation reactor includes a reactor pressure vessel, a cylindrical shroud disposed within the reactor pressure vessel to surround a reactor core and to extend to a position above the reactor core, and a coolant descending passage defined between the reactor pressure vessel and the shroud for allowing the coolant discharged from the top of the reactor core to be again supplied to the bottom of the reactor core. In such a natural circulation reactor, a liquid level of coolant in the coolant descending passage is adjusted on the basis of a reactor power request signal at a position below a liquid level of coolant in the shroud, and thus reactor power is controlled so that it may reach a predetermined reactor power by adjusting the liquid level of coolant in the coolant descending passage.

Method of and apparatus for controlling power of natural circulation

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. When the primary cooling water flows out, so that the pressure in a reactor vessel decreases, the emergency cooling water in the high-temperature region is boiled to expand to cause the low-temperature emergency cooling water in the lower portion of the water storage container to be raised, the resultant low-temperature emergency cooling water being supplied to the core region.

Nuclear reactor

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.

Emergency reactor core cooling structure

Growing community of inventors

Mito, Japan

Michio Murase