616results about "Nuclear plant auxillary equipment" patented technology

Method and system for supplying emergency power to nuclear power plant, wherein the method includes, providing accumulator battery system, connected to emergency bus, the accumulator battery system is monitored by online monitoring system; in case of power loss of electrical devices of the nuclear power plant, the online monitoring system starts the accumulator battery system to provide power supply to the electrical devices of the nuclear power plant via the emergency bus. The present application is adapt to the key technologies and battery management technologies of million kilowatt-class advanced pressurized water reactor nuclear power plant, facilitating to improve the safety of the nuclear power plant in case of serious natural disasters beyond design working conditions.

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.

A manipulator for servicing tubes extending through a tube sheet having three major components: a base member, a block member, and a foot member. The base member has a holder for supporting tooling or inspection devices. The base member also has at least one gripper for releasably gripping a tube extending through the tube sheet. The block member is connected to the base member for linear movement and rotation relative to the base member. The foot member is connected to the block member for linear movement relative to the block member. The foot member has at least one, and preferably two grippers, for releasably gripping a tube extending through the tube sheet. The manipulator supports itself using the two sets of expandable and translatable grippers that anchor into the tube ends and can move from tube to tube by alternately engaging each set of grippers.

The present invention relates to a nuclear power plant and safety system with fuse element and gravity elevator, the buildings of the power plant subjected to contamination being buried below sea level and under borated water basins, and having a safety system free of electrical and electronic components to act in the event of possible accidents comprising, among others, means for flooding the buildings of the power plant with thermal fuses and gravity elevators for operator evacuation in the event of an emergency.

The present invention relates to filters used to remove debris from water being sucked into a piping system. It has particular application 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. There are restrictions on the allowable pressure drop across the strainer and the space available for installing this equipment. The finned strainer of the present invention addresses these issues while maximizing the quantity of debris filtered from the water.

An apparatus that harnesses the thermal energy of spent fuel rods in nuclear power plants to power the cooling system of the nuclear power plant particularly the cooling for the spent fuel rod storage ponds and the main reactors. The apparatus is comprised of a heat exchanger unit that accumulates the thermal energy of the spent fuel rods, a heat conveyance system that conveys the thermal energy of the spent fuel rods, and a heat engine that receives its thermal energy input from the spent fuel rods and produces mechanical power that runs an electrical generator which powers the cooling system of the nuclear power plant, particularly the controls and pumps that cool the spent fuel rod storage ponds and the main reactors. The apparatus provides a redundant power source and makes the cooling system of nuclear power plants independent of externally supplied electrical power and thereby resolves a key redundancy and safety concern with nuclear power generation. The apparatus also has application to other industries.

The invention belongs to reactor design technology and particularly relates to a reactor cavity water injection cooling system with a combination of active and passive power. The reactor cavity water injection cooling system structurally comprises a passive reactor cavity water injection tank and reactor cavity water injection cooling pumps, the passive reactor cavity water injection tank is connected with a reactor cavity through a passive injection line, the reactor cavity water injection cooling pumps are arranged outside a containment, inlet tubes of the reactor cavity water injection cooling pumps are connected with a refueling water tank, and outlet tube lines of the reactor cavity water injection cooling pumps penetrate through the containment and are connected with the reactor cavity. As a countermeasure for serious accidents, the reactor cavity water injection cooling system has safety functions of taking away reactor core fusant, exhausting reactor core heat and avoiding melting in a multi-redundant and diversified way of combining active power with passive power when the accidents happen.

A ceramic heat exchanger includes a heat exchange section that heat-exchanges between two fluids A and B flowing opposite directions to each other. The heat exchange section includes ceramic blocks stacked one on top of another with a seal therebetween. The ceramic blocks have a plurality of parallel lines of flow channels, each line defined by the flow channels through which the same fluid flows, any two adjacent lines being defined by the flow channels through which the different fluids A and B flow respectively. Both ends in the stacking direction of the stack are bound to join and integrate the ceramic blocks with tightening means including end plates and a tie rod. A thermal expansion absorber is disposed on an external surface of the end plates for absorbing thermal expansion in the axial direction of the tie rod.

The invention provides a compactly arranged small-sized reactor primary loop overall structure. The structure is suitable for a marine operation environment, and comprises a pressure container, double-layer sleeve structures, inner-pile components, a control rod driving mechanism, a main pump, a steam generator, a pressure stabilizer, a main equipment integral support, a suppression pool, a primary shielding structure and the like. The steam generator is connected with the pressure container through one double-layer sleeve structure; the main pump is connected with the pressure container through one double-layer sleeve structure; the pressure stabilizer is connected with an evaporator through a surge pipeline; and the pressure container, the steam generator, the main pump and the pressurestabilizer are installed in the suppression pool through the integrated integral support. According to the system provided by the invention, the circulation of a nuclear reaction coolant in a reactorcoolant system can be ensured, realization of compact arrangement of a nuclear steam supply system is facilitated, and the heat exchange efficiency is improved.

The invention belongs to the reactor design technology, particularly relates to an active and non-active combined core water injection heat lead-out device which structurally comprises a safety injection pipe respectively connected with a refueling water tank and a heat pipe section and a cooling pipe section of a loop. A safety injection pump is arranged on the safety injection pipe, the safety injection pump comprises two independent medium-pressure safety injection pumps and two independent low-pressure safety injection pumps, the refueling water tank is arranged at the position of a pit below an inner core of a containment vessel, the safety injection pump is arranged on the outer portion of the containment vessel, a non-active safety injection tank and a core water compensating tank are further arranged above a reactor in the containment vessel, the safety injection tank is connected with the cooling pipe section of the loop through a pipe with a control valve, and the core water compensating tank is connected between the heat pipe section and the cooling pipe section of the loop through a pipe with a control valve. The active and non-active combined core water injection heat lead-out device enables active and non-active advantages to be combined, has the advantages of redundancy, diversity, high reliability and the like, and improves safety of a nuclear power plant.

An emergency temporary spent fuel pool cooling system for a nuclear power generating facility that has a permanently installed primary loop within the nuclear containment and a mobile temporary secondary loop. The secondary loop is housed in transport vehicles that can be stored off site and is connectable in heat exchange relationship with the primary loop through quick disconnect couplings that are accessible on the outside of the reactor containment. The transport vehicles also include self-contained power and compressed air sources for powering and controlling the entire emergency cooling system. The system also has a make-up water injection capability for refueling the spent fuel pool and secondary loop.

A nuclear power plant comprises a pressurized water reactor (PWR) and a steam generator driving a turbine driving an electric generator. A condenser condenses steam after flowing through the turbine. Responsive to a station blackout, the nuclear power plant is electrically isolated and a bypass valve is opened to convey bypass steam flow from the steam generator to the condenser without flowing through the turbine. The thermal power output of the PWR is gradually reduced over the transition time interval. After opening, the bypass valve is gradually closed over the transition time interval. A supplemental bypass valve may also be opened responsive to the station blackout to convey supplemental bypass steam flow from the steam generator to a feedwater system supplying secondary coolant feedwater to the steam generator. The supplemental bypass steam flow does not flow through the turbine and does not flow through the condenser.

Liquid nitrogen, the densest and highly transportable form of nitrogen, and the cold nitrogen gas it produces when released from its container, for emergency cooling of fuel rods and nuclear reaction chambers and fighting hazardous fires is disclosed. Liquid nitrogen may be disbursed via a thermally-activated heat valve attached to a container of liquid nitrogen.

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 quick disconnect for a control rod drive mechanism seismic support tie rod system that is remotely operable from a nuclear power plant's operating deck. A wall mounted anchor in the reactor cavity contains one half of a disconnect coupling that interfaces with the other half of the disconnect coupling on the ends of the tie rods employing a remote winching system that is actuated from the top of the reactor head assembly. A latching mechanism is then actuated from the refueling cavity operating deck to lock the tie rod in place and prevent displacement during a seismic or pipe break event. The tie rod may similarly be unlocked from the wall anchor and raised above the reactor head assembly as part of a reactor head disassembly operation to gain access to the core of the reactor vessel for refueling.

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 in the fluid to the cooling water storage unit.

A nuclear engineering plant has a containment, whose interior chamber is subdivided by a wall into a systems chamber and an operating chamber which is accessible during normal operation. The containment ensures a particularly high operational reliability, in particular also in incident situations, in which hydrogen is released in the systems chamber. For this purpose, a number of overflow openings are provided in the partition wall, the respective overflow opening is closed by a closure element of a closure apparatus which opens automatically when a trigger condition associated with the respective overflow opening is reached. Closure apparatuses are provided which open both as a function of pressure and independently of pressure. The closure apparatus furthermore has a closure element containing a bursting film or a bursting diaphragm. The closure apparatus is configured such that it frees the overflow opening automatically when a predetermined environment-side trigger temperature is reached.

The invention relates to a special transportation device of a nuclear power station for a narrow space, comprising a mounting plate and a chassis, wherein the bottom of the mounting plate is rotatably connected with the chassis, and a positioning telescopic arm is arranged between the mounting plate and the chassis; and baffle plates are arranged at the bottom and/or two sides of the mounting plate. As the mounting plate and the chassis are rotatably connected, the inclination angle of the transportation device can be adjusted so as to put an equipment cabinet onto the mounting plate obliquely, thereby the transportation can be ensured to be convenient, and the equipment cabinet can be prevented from generating the phenomena of downward sliding and lateral sliding in the transportation process.

A degassing plant (2) for reactor coolants (R) has a degasser column (6) which is coupled to the primary cooling circuit of a pressurized water reactor and through which reactor coolant (R) flows. The degassing plant (2) furthermore comprises a coolant evaporator (54) with a first heat exchanger (52) and a strip steam condenser (34) with a second heat exchanger (44), wherein a partial flow of thereactor coolant (R) flows through the heat exchanger (52) of the coolant evaporator (54) on the secondary side, and wherein the heat exchanger (44) of the strip steam condenser (34) is switched, on the primary side, in a vapour and gas outlet line (32) which is connected to the degasser column (6). The degassing plant (2) is intended to be configured such that, with as simple a design as possibleand taking into consideration relevant safety procedures, a particularly effective and at the same time energy-efficient separation of gasses, which are dissolved in the reactor coolant (R) and cannot be condensed, is made possible, wherein the thermal load of the assigned nuclear intermediate cooling system is furthermore intended to be kept as low as possible. To this end it is provided that the heat exchanger (52) of the coolant evaporator (54) is switched in a heat-pump circuit (58) on the primary side, which heat-pump circuit is coupled to the heat exchanger (44) of the strip steam condenser (34) with respect to the heat flux, which is established during plant operation, such that the heat liberated in the strip steam condensation is transferred at least partially to the reactor coolant (R), which flows through the coolant evaporator (54), and thus leads to its evaporation.

The invention discloses an underground nuclear plant with nuclear island cavern groups distributed in an L shape. The underground nuclear plant comprises nuclear islands used for generating steam through nuclear fission energy and arranged in underground cavern groups, a reactor factory building cavern used for containing reactor factory buildings is arranged in the middle in the underground cavern groups, and auxiliary cavern groups used for containing auxiliary factory buildings are arranged around the reactor factory building cavern in an L shape. By means of the underground nuclear plant with the nuclear island cavern groups distributed in the L shape, an existing underground cavern digging technology meets the building requirement of the underground nuclear plant, the number of caverns contained in the auxiliary cavern groups is small, and stability of surrounding rock is facilitated.

A device for accumulating, isolating, indicating and venting accumulated gas in a fluid system pipe includes a main pipe fitting affixed to a system pipe in which a hole has been drilled. A stand pipe attached to the pipe fitting houses a magnetic float. A magnetic-float level indicator exterior to the pipe and responsive to the magnetic float indicates the magnetic float's level. A valve attached to the stand pipe above the magnetic float allows controlled ventilation of the gas in the standpipe and thus in the piping system. Gas from the system pipe accumulates in the standpipe, removed from the primary fluid flow path of the system pipe. In the standpipe, as the liquid/gas interface drops, the float drops to a pre-determined level, at which point the user vents the gas from the piping system, causing the magnetic float to rise to a level indicating that gas is again at acceptable levels in the piping system.

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.

The invention discloses a nuclear power station conventional island desalted water distribution system full-stop overhauling method. The method comprises the steps that A, ice blockage is carried out on a main pipeline between a conventional island desalted water distribution system and a unit, and an isolating device and a connecting device are additionally arranged at the downstream part of ice blockage; B, a temporary water supply pipeline is connected to the connecting device from an SED, operation verification is carried out on the temporary water supply pipeline, and the temporary water supply pipeline is communicated with the main pipeline between the conventional island desalted water distribution system and the unit through the connecting device after the temporary water supply pipeline is qualified; C, the connecting device is opened to supply water to the unit, and then the isolating device is closed; D, the conventional island desalted water distribution system is made to stop fully, isolating and emptying are carried out, and damaged or aged devices on the main pipeline are replaced; E, a drain valve and a vent valve are opened fully, the system is filled with water, air in the system is exhausted, the system is washed, samples are taken for a laboratory test, and after the samples are qualified, the drain valve is closed, and the conventional island desalted water distribution system is put into operation; F, the connecting device is disconnected with the temporary water supply pipeline, and the connecting device is sealed. The nuclear power station conventional island desalted water distribution system full-stop overhauling method is convenient and rapid to implement, safe and low in cost.

A nuclear power plant and an operation method of the nuclear power plant, which can uprate plant power without greatly modifying the construction of plant equipment, while keeping a core's pressure loss characteristic, a safety margin, and a design margin in the transient state substantially the same as those before the power uprate. Thermal power in a second operation cycle of a nuclear reactor is uprated from thermal power in a first operation cycle preceding the second operation cycle at least one operation cycle. A proportion of steam extracted from a steam system and introduced to a feedwater heater, which is in particular extracted from an intermediate point and an outlet of a high pressure turbine, with respect to a flow rate of main steam is reduced in the second operation cycle from that in the first operation cycle such that temperature of feedwater discharged from the feedwater heater lowers in the range of 1° C. to 40° C. in the second operation cycle.