103results about "Plant pumping arrangements" patented technology

An auxiliary system for cooling a spent nuclear fuel pool through a submersible heat exchanger to be located within the pool. In each train or installation, a single loop or series of loops of cooling fluid (e.g., sea water or service water) is circulated. The system is modular, readily and easily installed during an emergency and can be self operating with its own power source. Multiple trains may be used in parallel in order to accomplish the required degree of spent fuel pool cooling required.

The present invention relates to filters used to remove debris from water being sucked into a piping system. It has particular application for use in nuclear power plants, which, after a loss of coolant accident, must pump cooling water back into the reactor core from a collection sump. This water may contain various types of debris that must be removed before the water is sent back into the reactor cooling system. Filtering of the debris is realized with the component known as “strainers”. There are restrictions on the space available for installing strainers. The vaned filtering element, for example a vaned fin, of the present invention is designed to reduce the space required for strainer installation by increasing strainer surface area per unit volume, while maximizing the quantity of debris that can be filtered from the water.

The present invention discloses a nuclear reactor coolant pump that does not rely on an electric motor, but is operated by means of driving force generated inside a nuclear power plant, so a to be capable of maintaining the safety of the nuclear reactor when the nuclear reactor is operating normally and also in the event of an accident in the nuclear reactor. The nuclear reactor coolant pump comprises: a pump impeller rotatably installed in a first fluid passage of a nuclear reactor coolant system to circulate a first fluid inside the nuclear reactor coolant system; a drive unit receiving steam from a steam generator to generate driving force to rotate the pump impeller, and rotating about the same rotating shaft as the pump impeller to transfer the generated driving force to the pump impeller; and a steam supplying unit forming a passage between the steam generator and the drive unit to supply at least a portion of the steam released from the steam generator to the drive unit.

Disclosed is a nuclear power plant which drives a Stirling engine by means of heat generated in nuclear power plant safety systems during an accident, uses the resulting power directly or generates electric power so as to supply the power to the safety systems, and thus can improve economic efficiency as well as the reliability of safety systems, such as a passive safety system, by operating the safety systems without an emergency diesel generator or external electric power.

An apparatus for reinforcing a jet pump riser includes: an elbow upper clamp for covering a riser elbow coupled to a thermal sleeve from an upper side thereof; an elbow lower clamping member for clamping the riser elbow from a lower side thereof; an elbow vertical portion clamping member for covering a vertical portion of the riser elbow; and an elbow horizontal portion clamping member for covering a horizontal portion of the riser elbow. These members are disposed in different orientations with respect to the elbow upper clamp so as to fix the thermal sleeve, the riser elbow and the riser pipe.

The invention belongs to the pressurized water reactor overall structure design technology, and specifically discloses an integrated small reactor. The integrated small reactor comprises a pressure vessel, and a reactor core, a compressing assembly, a control rod assembly, a guide assembly, a control rod driving mechanism and a once-through steam generator which are arranged inside the pressure vessel, and further comprises a main pump directly connected with a cylindrical body of the pressure vessel, wherein the reactor core comprises a reactor core barrel and a fuel assembly; the compressing assembly and the control rod assembly are arranged inside the reactor core barrel; and the pressure vessel is directly connected with the main pump, the main pump is horizontally installed, and an inlet thereof is provided with a flow distribution device. Heating equipment is arranged on a top cover of the pressure vessel, a coolant inside the pressure vessel is subjected to pressure control directly through temperature adjustment, that is, main equipment such as the steam generator, voltage-stabilizing equipment and the control rod driving mechanism are arranged inside the pressure vessel of the reactor at the same time. According to the integrated small reactor, the external loop pipeline is omitted, large break accidents are eliminated, and the structure is compact; and the control rod driving mechanism is arranged internally, the rod ejection accident is eliminated, and the fixation safety is greatly improved.

An apparatus for cooling a spent fuel pool having a heat exchanger includes a cooling water pool positioned above the spent fuel pool; a floating device configured to be elevated according to a water level of a cooling water in the spent fuel pool; and an emergency cooling water supply pipe configured to form a path through which the cooling water of the cooling water pool is moved to the spent fuel pool and configured to include a floating valve that opens or closes a flow passage of the cooling water in connection with the elevation of the floating device.

The invention relates to structural design of an emergency fresh air pipeline of a ventilation system of a main control room of a nuclear power station. The emergency fresh air pipeline comprises two accident air taking ports at different positions of a plant roof of the nuclear power station, and the two accident air taking ports are opened or closed through isolating valves respectively; the two accident air taking ports are connected by an air pipe and share one identical emergency air sending pipeline and are connected with an emergency accident air blower, the emergency accident air blower introduces fresh air from the outdoor into the main control room, and an iodine absorption device is arranged on the emergency air sending pipeline; and the two accident air taking ports are arranged in opposite directions relative to a reactor containment. The emergency fresh air pipeline disclosed by the invention reduces the risk of losing habitability of the main control room under the working condition of the accident, and improves the reliability and safety of the ventilation system of the main control room.

The invention provides a nuclear power station primary pump of the type comprising a drive shaft for rotating a wheel, a heat exchanger, a thermal barrier cover comprising a peripheral skirt provided at its base with a circular end wall, and a diffuser comprising a peripheral skirt provided at its base with a rim. The circular end wall co-operates with the base of the heat exchanger to define a bottom chamber for cold water and includes a central bore provided with a succession of circumferential and superposed rings forming a lip seal of the labyrinth type, and a space is formed between the thermal barrier cover and the diffuser to maintain stagnant cooling fluid.

A packing seal is provided for a system for sealing the shaft of a primary motor-driven pump unit of a nuclear reactor, intended to ensure sealing between the primary circuit and the atmosphere. The packing seal includes a rotary active surface and a floating active surface, and a face of the floating active surface and / or the rotary active surface is covered by a protective layer made from a material having surface energy greater than 30 mJ / m2 and an electron donor component less than 15 mJ / m2.

The invention discloses a shaft sealing water injection system of a main pump of a nuclear power station. The shaft sealing water injection system comprises an injection pump, a high-pressure cooler, a hydrocyclone, valves and a main connecting pipeline that are arranged at the exterior of the main pump, as well as an auxiliary pump and an internal runner that arranged in the main pump; the internal runner of the main pump is connected with an external runner of the main pump through a shaft sealing water injection hole and a high temperature water drainage hole that are formed in a main flange; and the injection pump, the high-pressure cooler and the hydrocyclone are sequentially arranged on the main connecting pipeline. The shaft sealing water injection system further comprises a bypass pipeline connected with the injection pump, the high-pressure cooler and the hydrocyclone in parallel; the main connecting pipeline is provided with a normally open main pipeline isolating valve at each of the two ends of a parallel connection section connected with the bypass pipeline in parallel; and the bypass pipeline allows low temperature upward filling water of a RCV system to be able to bypass the injection pump, the high-pressure cooler and the hydrocyclone and to directly enter the shaft sealing water injection hole of the main flange. Compared with the prior art, through redundancy design, and optimization of a system structure and control, the safety and the reliability of system operation and the maintainability of the shaft sealing water injection system are ensured.

A method and apparatus for providing an alternative remote spent fuel pool cooling system for the spent fuel pool. The cooling system is operated to cool the spent fuel pool in the event of a plant accident when normal plant electricity is not available for the conventional fuel pool cooling and cleanup system, or when the integrity of the spent fuel has been jeopardized. The cooling system is operated and controlled from a remote location, which is ideal during a plant emergency.

The invention belongs to the technical field of primary coolant circuit auxiliary systems of million kilowatt-class nuclear power stations, and provides a purging method of a pressurized-water reactornuclear island vessel. A hydrogen discharging purging system for purging of the pressurized-water reactor nuclear island vessel is included. In the processes of water replenishment and nitrogen purging of a pressure-relief box, a check valve bank is arranged on a second pipeline allowing a nuclear island nitrogen supply system to be connected with the pressure-relief box, the back pressure in thepressure-relief box is lowered, the pressure-relief box can be filled with water through a water replenishment system, and the space of hydrogen-containing waste gas in the pressure-relief box is allcompressed and the hydrogen-containing waste gas is all discharged from the pressure-relief box; then the pressure-relief box is stopped replenishing and draining water until the water level in the pressure-relief box drops to the normal water level, the nuclear island nitrogen supply system replenishes nitrogen into the pressure-relief box through the second pipeline, the space available for gasstorage in the pressure-relief box is quickly filled, the hydrogen-containing waste gas in the pressure-relief box is all purged and discharged, the hydrogen content in the pressure-relief box is lowered to 0.5 V / V% or below, and the risk of hydrogen explosion is avoided.

The invention discloses a drainage system and a drainage method for an underground nuclear power plant cavern group. The drainage method comprises the steps that accumulated water in an underground nuclear power plant cavern flowing out via a drainage hopper is monitored by means of a radioactivity monitor, a controller shuts off a first isolation valve when the radioactivity monitor monitors that the accumulated water is radioactive, the controller opens a second isolation valve, the radioactive accumulated water enters a drainage branch associated with radioactivity, and is discharged into a siphon well after being treated by a radioactive substances treatment room; when the radioactivity monitor does not monitor that the accumulated water is radioactive, the controller shuts off the second isolation valve and opens the first isolation valve, and the non-radioactive accumulated water is discharged into the siphon well; and the accumulated water is discharged to rivers outside a mounting by means of the siphon well. According to the drainage system and the drainage method, the passive drainage system is constructed by utilizing a siphon pipeline and the siphon well, negative pressure formed in a siphon device forces water in the pipeline to flow under pressure, the flow velocity is accelerated, the drainage effect is good, and the drainage system can discharge the accumulated water out of the underground cavern in an organized and efficient manner regardless of accidental water accumulation or serious accident conditions.

The invention relates to a passive advanced pressurized water reactor coolant pump installation technology. The passive advanced pressurized water reactor coolant pump installation technology is characterized in that a main pump occasional passage is provided; before a main pump is introduced to a nuclear island workshop, butt joint between a flat car transporting the main pump and the main pump occasional passage is completed; a draw gear is used to introduce a main pump / transport towing bracket assembly embarked on a mobile device to the nuclear island workshop; during the drawing process, a tail slipping device is used to prevent the moving speed becoming too fast, and then the main pump is moved to a turnover initial position; a ring crane hook is lifted and coordinates with moving of the mobile device so as to turnover the main pump, and at the same time, a tow rope of the tail slipping device is adjusted; the main pump is lifted to a steam generator cavity, and is located in a main pump installation mechanical device; a suction adapter in a pump housing completes the installation work before installation of the main pump; the main pump installation mechanical device is utilized to provide moving, positioning, adjusting and lifting of the main pump; and a sealing ring at the upper part of the main pump and the pump housing directly utilize the welded sealing technology to realize prevention of leakage during the operating process.

Systems and methods for upgrading power output of previously-deployed nuclear power plants are described. Systems and methods may include a base nuclear power plant with a predetermined base power output rating and a predetermined base whole core refueling interval. Systems and methods may also include a power upgrade kit for increasing the base power output rating from the base power output rating to an increased power output rating without a change in fuel charge, reactor structures, or civil structures.

The invention discloses a novel mobile emergency diesel driven water injection device for a nuclear power station. The novel mobile emergency diesel driven water injection device comprises a water injection pump unit, a drive device used for driving the water injection pump unit, and a mobile load device used for bearing the pump unit and the drive device, wherein the drive device is a diesel engine, an output shaft of the drive device is connected with an input shaft of the water injection pump unit through a clutch, the clutch is used for controlling whether the drive device provides power to the water injection pump unit; the drive device and the water injection pump unit are arranged on the mobile load device. According to the invention adopting the technical scheme, the drive device drives the water injection pump unit to inject water, and a targeted monitoring device and a targeted regulating device are arranged in a pipeline of the water injection pump unit for ensuring that the water injection amount is controlled in real time; the novel mobile emergency diesel driven water injection device is installed on the mobile load device, thus the maneuverability of the novel mobile emergency diesel driven water injection device is ensured.

The invention belongs to the technical field of nuclear reactors. The invention particularly relates to a heat pipe reactor system based on thermophotovoltaic power generation. The system comprises aheat pipe reactor power supply main system, a thermophotovoltaic generator cooling system, an in-cabin pressure maintaining system, an emergency reactor core cooling system, a waste heat leading-out system, a temperature and humidity control system and an integral pressure-resistant shell, all the mentioned units forming a modular device; the heat pipe reactor power supply main system is positioned at the middle part in the integral pressure-resistant shell; the thermophotovoltaic generator cooling systems are symmetrically arranged on two sides of the heat pipe reactor power supply main system; the thermophotovoltaic generator cooling systems are positioned on the integral pressure-resistant shell; the cabin pressure maintaining system is positioned at the bottom of the tail part of the integral pressure-resistant shell; the temperature and humidity control system is positioned on the top wall of the tail part of the integral pressure-resistant shell; and the waste heat leading-out system is positioned outside the bottom wall of the integral pressure-resistant shell. The system disclosed by the invention has the advantages of long service life or even full service life, no material change, high inherent safety, low noise, high power volume-weight ratio and simple and reliable system equipment.

A system for injecting hydrogen into Boiling Water Reactor (BWR) reactor support systems in operation during reactor startup and / or shutdown. The system the hydrogen injection system includes at least one hydrogen source, flow control equipment, and pressure control equipment. The pressure control equipment being configured to regulate a pressure of a hydrogen flow between the at least one hydrogen source and the at least one first BWR support system based upon an operating pressure of the at least one first BWR support system.

Object of this invention is to provide an axial torsional vibration measurement device of the reactor internal pump capable of measuring the axial torsional vibration easily. The vibration detector 2 that detects vibration in the radial direction, and the sound detector 3 that detects vibration in the circumferential direction are fitted on cylindrical motor casing 6, which contains the motor of a reactor internal pump. The dimension of both signals of a vibration waveform signal detected by the vibration detector 2, and a sound waveform signal detected by the sound detector 3 are matched in vibration calculating device 15, to thereby obtain an axial torsion vibration of a motor from the difference of both signals.

A nuclear power plant comprising a primary coolant circuit, a steam-water circuit separated from the primary coolant circuit and a steam generator connected to the primary coolant circuit and the steam-water circuit to transfer heat from the primary coolant circuit into the steam-water circuit has at least one dosing point in the steam-water circuit to inject a reducing agent into the steam-water circuit, wherein the reducing agent is an organic compound consisting of carbon, hydrogen and oxygen. Furthermore, a method for operating said nuclear power plant and the use of the method for downtime preservation of the secondary side of a steam generator of a nuclear power plant are provided.

Disclosed is a nuclear power plant which drives a Stirling engine by means of heat generated in nuclear power plant safety systems during an accident, uses the resulting power directly or generates electric power so as to supply the power to the safety systems, and thus can improve economic efficiency as well as the reliability of safety systems, such as a passive safety system, by operating the safety systems without an emergency diesel generator or external electric power.

A pump unit for a nuclear installation, for example a pressurized water reactor, contains a vertical shaft, which is guided through at least one journal bearing. The journal bearing or bearings is or are configured as a spiral groove bearing or bearings, to prevent shaft vibration caused by the bearings and to increase the service life of the pump unit. Preferred uses of the pump units include, for example, the coolant pump of a pressurized water reactor and the circulating pump of a boiling water reactor.

The invention discloses a primary circuit coolant charging and discharging system for a nuclear power plant reactor. The primary circuit coolant charging and discharging system comprises a primary circuit pressurization subsystem used for charging high-temperature gas into a primary circuit reactor main vessel for pressurization when a coolant is discharged into a primary circuit, and a coolant delivery subsystem which comprises a primary circuit coolant charging circuit, a primary circuit coolant discharging circuit, a coolant delivery pump and a coolant storage device, wherein the coolant storage device is connected with the primary circuit reactor main vessel through the primary circuit coolant charging circuit and the primary circuit coolant discharging circuit; the coolant delivery pump is arranged in the primary circuit coolant discharging circuit; the primary circuit coolant charging circuit is used for charging a coolant into the primary circuit reactor main vessel; and the primary circuit coolant discharging circuit is used for pumping the coolant in the primary circuit reactor main vessel to the coolant storage device. According to the invention, through cooperation of primary circuit gas phase pressurization with discharging of a primary circuit high-density coolant pumped by the coolant delivery pump, the coolant is pressurized to be discharged, so devices in the main vessel can be easily maintained and replaced.

A system and a method for injecting hydrogen into Boiling Water Reactor (BWR) reactor support systems in operation during reactor startup and / or shutdown to mitigate Inter-Granular Stress Corrosion Cracking (IGSCC). The system may provide hydrogen at variable pressures (including relatively higher pressures) that match changing operating pressures of the reactor supports systems as the reactor cycles through startup and shutdown modes.

The present invention relates to filters used to remove debris from water being sucked into a piping system. It has particular application for use in nuclear power plants, which, after a loss of coolant accident, must pump cooling water back into the reactor core from a collection sump. This water may contain various types of debris that must be removed before the water is sent back into the reactor cooling system. Filtering of the debris is realized with the component known as “strainers”. There are restrictions on the space available for installing strainers. The vaned filtering element, for example a vaned fin, of the present invention is designed to reduce the space required for strainer installation by increasing strainer surface area per unit volume, while maximizing the quantity of debris that can be filtered from the water.