8983 results about "Nuclear power" patented technology

A nuclear-powered plant for systems of up to about 100 MWs with a confinement section where the reaction takes place in a core having a reactive thorium / uranium-233 composition, and where an external neutron source is used as a modulated neutron multiplier for the reactor core output. The core is housed in a containment structure that radiates thermal energy captured in a multiple-paths heat exchanger. The exchanger heat energy output is put to use in a conventional gas-to-water heat exchanger to produce commercial quality steam.

The invention relates to a silicon dioxide aerogel material and a preparation method thereof. The method comprises the following steps: by using water glass as silicon source, adding an acid-containing organic solvent free of chlorine ions and fluorine ions to generate a precipitate of sodium ions, potassium ions and other metal salt ions, filtering to remove the precipitate to obtain high-purity silica sol, carrying out a sol-gel process, aging, acidifying, modifying, and drying to obtain the silicon dioxide aerogel material. The acidification before modification enhances the surface reaction activity of the silicon gel, thereby obviously enhancing the modification effect and efficiency. The method has the advantages of low cost and simple and efficient technique, is beneficial to mass high-efficiency production, is free of chlorine ions and fluorine ions in the whole technical process, and enhances the equipment operation safety and reliability; and the product can be used for heat preservation and thermal insulation of nuclear power and liquefied natural gas equipment and pipelines with higher requirement for corrosion resistance, and can also be used for thermal insulation in the field of aerospace, petrochemical engineering, track transportation, ships, automobiles, construction and the like.

The invention relates to a nuclear power plant working robot and a control system thereof, which belong to the fields of robots and automation equipment. The nuclear power plant working robot is a crawler-type mobile manipulator and is composed of a mobile platform driven by double crawlers and a four freedom degree manipulator carried on the mobile platform. The nuclear power plant working robot can move inside a nuclear power plant, has two control modes of manual remote control and autonomous control, and is remotely controlled by a wireless or wired mode. The control system of the nuclear power plant working robot includes a host monitoring and planning control system and a robot control system which are matched to control the operation of the robot. The robot operates autonomously, is safe and reliable, can complete certain dangerous tasks in high radioactivity environment; the robot has small size, stable performance, great maneuverability and low operation cost; a crawler-type chassis has strong gripping force as well as certain climbing and obstacle clearing ability, and is suitable to walking over complex terrains; and the robot has high intelligence degree and can realize autonomous control and manual remote control.

Methods are disclosed for nondestructive testing and measuring the structural health of prestressed concrete structures, such as slabs, columns, girders, bridges, towers, elevated storage tanks, silos, cooling towers, wind power generation towers, liquefied gas storage tanks, nuclear power containment buildings, and the like. Measurements are made as the structure undergoes tensioning and detensioning operations. By measuring actual movements of cardinal points on the structure, in an absolute three-dimensional coordinate system, and comparing the measurements to a model—as tension on a tendon is changed—a margin of safety is assured. High accuracy measurements are made by electronic distance measurement (EDM) instruments over hundreds of meters, which yield coordinates of cardinal points with an uncertainty of the order of one part per million. The methods are proposed as possible alternatives to prior failures of post-tensioned concrete, including the Las Lomas Bridge, the Kapiolani Interchange On-Ramp, Turkey Point Unit 3 Nuclear Power Plant, and Crystal River Unit 3 Nuclear Power Plant. An extensive review of the most closely related prior arts is included.

The invention belongs to safety equipment for a nuclear power plant, in particular to a passive and active combined special safety system for the nuclear power plant. The system comprises a passive high-pressure core water supply tank, an accumulator, a low-pressure safety injection pump, a refueling water tank, a secondary-side passive residual heat removal heat exchanger, a vapor condensation water tank, a containment spray system, a passive reactor cavity water injection system, related valves and related pipes. When the nuclear power plant has a design basis accident or beyond design basis accident, through a series of passive and active safety facilities used in all steps, the circuit of a reactor and the core can be cooled effectively, so that the nuclear power plant can enter a safe cold shutdown condition smoothly; and thus, the serious accident consequence of the reactor is inhibited or relieved, damage caused by accidents is relieved, and the safety of the nuclear power plant is improved.

The invention discloses an automatic vehicle navigation method, a navigation pattern information compiling method and vehicle navigation equipment. The automatic vehicle navigation method includes: displaying navigation pattern information on a road; acquiring the navigation pattern information through the vehicle navigation equipment; and regulating vehicle running parameters by the vehicle navigation equipment according to the acquired navigation pattern information. By the automatic vehicle navigation method, a function of satellite-free automatic vehicle driving can be realized, and automatic vehicle navigation for culverts, tunnels and dense high-rise building districts can be realized. The ground navigation pattern information acts in a close range, so that accuracy is high. Implementation cost of the automatic vehicle navigation method is low, so that whole-system implementation cost in automatic vehicle navigation can be lowered effectively. In addition, automatic power transmission, automatic circuit switching for vehicle lane changing and the like can be achieved for electric vehicles during automatic navigation, so that wide-range application of electric vehicles can be availed, dependence of traffic systems on petroleum is lowered and application of wind power and nuclear power in road traffic systems is achieved.

The invention discloses a nuclear power plant non-security-level DCS configuration testing method and system comprising a virtual DCS platform and a nuclear power station simulating model coupled to the virtual DCS platform. The nuclear power station simulating model is used for, according to a nuclear power unit operation control instruction given by the virtual DCS platform, simulating the operation of a real nuclear power station, acquiring the operating data of the nuclear power unit, and returning the acquired data to the virtual DCS platform. The virtual DCS platform is used for, according to the operating data of the nuclear power unit, computing and analyzing the operating condition of the nuclear power unit in real time, evaluating a difference between an analyzed result and a designed operating condition, determining the correctness and the rationality of DCS configuration logic and parameters, giving corresponding logic modification and parameter optimization, and transplanting and applying a tested and modified result to the DCS configuration of a real nuclear power unit. The nuclear power plant non-security-level DCS configuration testing method and system may reduce the investment cost of the nuclear power station, shorten the construction period of nuclear power engineering, and decrease equipment loss.

The invention relates to a microminiature operation underwater robot of a nuclear power plant, and the robot comprises an underwater robot body and an onshore control system, wherein two propellers are respectively arranged in the horizontal and vertical directions of the robot body, a depth gauge is arranged on the right side of the front part of the robot body, a manipulator is arranged at the bottom in front of the robot body, and a sonar is arranged on the top of the robot body; a control cabin is arranged in the middle rear part of the robot body; an outer cabin is sealed by a transparent glass cover, and is provided with a rearview camera and an auxiliary lighting light-emitting diode (LED) lamp; a front-view camera system is arranged at the front part of the robot body, and comprises a zooming radiation resistant camera tube, a tripod head and a lighting lamp; a video image and control signal is transmitted to the onshore control system through a shield cable; and the control system comprises a movement control rod, a manipulator control button, a keyboard, a main display screen and a speed governing knob. The microminiature operation underwater robot of the nuclear power plant is used for the monitoring and the simple foreign body fishing of a core pool, a spent fuel pool and a component pool of the nuclear power plant.

The invention relates to an active-passive combined reactor core residual heat removal system for a nuclear power station. The active-passive combined reactor core residual heat removal system comprises a safety injection system, a containment spraying system, an auxiliary water supply system, a reactor cavity water injection system, a secondary side passive residual heat removal system, a passive containment heat leading-out system and related valves and pipelines. The active-passive combined reactor core residual heat removal system provided by the invention realizes three safety functions of controlling reactivity, discharging reactor core heat and containing radioactive substance when an accident occurs in an active-passive combined multi-redundancy diversity manner. The active-passive combined reactor core residual heat removal system provided by the invention can completely realize the safety functions of safety injection and safety spraying under the condition that the accident occurs in a nuclear power plant, reactor cavity water injection under the serious accident working condition and the like, effectively improve reliability of a safety system, enhance coping capability of the safety system under the working condition that the accident occurs in the nuclear power station, can effectively prevent and relieve the serious accident, reduce the reactor core melting probability and risk probability of large-scale radioactivity release and greatly improve safety performance of the nuclear power station.

The invention relates to a method and system for improving reliability of emergency power supplies of a nuclear power plant. At least one path of high-capacity battery energy storage system (fixed energy storage system and / or movable energy storage system) is used to supplement or replace final emergency power supplies of a nuclear power plant. When all existing emergency power supplies of the nuclear power plant fail, the system is started to supply power for equipment in an emergency plant of the nuclear power plant so as to keep discharging waste heat of reactor cores and cooling spent fuel pools of the nuclear power plant, thus ensuring the safety of the nuclear power station. The method and system provided by the invention can be used for resisting against extreme weathers and avoiding common mode failure of the emergency power supplies in the extreme weathers, can strength the reliability of the emergency power supplies of the nuclear power plant and can at least reduce the melting probability of reactor cores by 21.6%, thereby improving the integral nuclear safety level of the nuclear power plant.

The invention discloses a modeling method for a third generation pressurized water reactor nuclear power generating unit. The modeling method comprises the following steps of: 1, decomposing a nuclear power generating unit system into a plurality of subsystem models; 2, establishing the subsystem models in the step 1 according to heat engineering and energy transfer and conversion rules; 3, combining the subsystem models obtained in the step 2 into a nuclear power generating unit full system model, and connecting the nuclear power generating unit full system model with a power system model toobtain a combined model of a nuclear power generating unit and the power system; and 4, establishing a customized model of the third generation pressurized water reactor nuclear power generating uniton the basis of the combined model of the nuclear power generating unit and the power system, and simulating the performance of the nuclear power generating unit and machine-grid interaction according to the customized model. The method effectively solves combined emulation of the nuclear power generating unit and a power unit, can be applied to machine-grid coordination analysis of a nuclear power plant and a power grid, and has high practicability.

This invention relates to a method and apparatus for nuclear power station equipment risk evaluation by computer. Wherein, using a computer to quantize the fault effect and store the information in a database; creating a logic relation between fault and facilities abnormity, as well as the abnormity and parameters; periodic acquiring device real-time data to substitute into the logic relation and calculate probable fault and risk; if over the early-warning value, producing alarm signal. This invention can monitor the plant for security, automatic evaluates risk to eliminate human factor, and obtains much accurate result.

The invention discloses a method for evaluating risk of nuclear power station, comprising the following steps: receiving the maintenance device state value transmitted by the user device, wherein the maintenance device is a power plant device to be regularly maintained by the user or regularly tested or a power plant device for evaluating risk value when something is wrong with the device, computing the risk evaluation value of the nuclear power station based on the risk real-time evaluation model according to the maintenance device state value, and transmitting the computed risk evaluation value to the user device where the user is. The invention further discloses a device for evaluating risk in real time. The embodiment of the invention is convenient for the operator to realize the riskcondition of the nuclear power station in real time.

A method for electrical cable testing by Pulse-Arrested Spark Discharge (PASD) uses the cable response to a short-duration high-voltage incident pulse to determine the location of an electrical breakdown that occurs at a defect site in the cable. The apparatus for cable testing by PASD includes a pulser for generating the short-duration high-voltage incident pulse, at least one diagnostic sensor to detect the incident pulse and the breakdown-induced reflected and / or transmitted pulses propagating from the electrical breakdown at the defect site, and a transient recorder to record the cable response. The method and apparatus are particularly useful to determine the location of defect sites in critical but inaccessible electrical cabling systems in aging aircraft, ships, nuclear power plants, and industrial complexes.

The invention belongs to the technical field of pipe whip restraint steel structures of main power houses of nuclear power station conventional islands, in particular to a production process method of the pipe whip restraint steel structure of the main power house of the nuclear power station conventional island, which comprises the lofting process, the marking-off and scribing process, the cutting processing process, the bevel processing process, the hole-making process, the correction process, the polishing surface, the material splicing process, the pre-splicing process of a steel member, the assembly process, the processing process of a friction surface, the welding process of the steel structure and the anti-slip test of the friction surface of the steel member before leaving a factory, sandblasting, derusting and coating of a base coat are carried out in the factory, and the production of the steel structure is divided into the production processing of a welded H-shaped steel column, the production of a steel beam, the assembly and the welding of a support part, the production of a combined steel beam, the assembly and the welding of a girder box body and an inclined support below the girder box body, and the assembly and the welding of a pipe whip restraint key. The production process method can improve the drilling precision, utilize a numerical control planar drilling machine to complete the drilling on a gusset plate, determine a variety of processing processes and welding process parameters, and ensure the dimension of the member after the assembly and the smooth proceeding of field installation.

The invention relates to an installing method for a main pipeline of a coolant system of a nuclear power station, which is characterized in that the main pipeline of the coolant system of a reactor of the pressurized water reactor nuclear power station comprises a cool section (41) and a hot section (42), and a steam generator is directly connected with a main pump. According to the invention, the problem that the main pipeline is installed and welded by only a pressure container or steam generator in place by adopting the installing method of the main pipeline without a transition section issolved, and technical limit that the main pipeline is installed after the pressure container is in place and the steam generator or main pump is in place in the traditional main pipeline constructiontechnology is eliminated. A main pipeline groove is processed by adopting a site numerical control machining technology, the main pipeline and equipment connected with the main pipeline are subjectedto measurement, modeling and process monitoring by adopting a laser tracking measuring and 3D modeling technology, the main pipeline is regulated to meet the assembly welding requirement and is installed by using a narrow TIG (argon tungsten-arc welding) automatic welding technology, thus a feasible method is provided for shortening the construction period of the nuclear power station.

The invention relates to a method for monitoring pressure pipe leakage acoustic emission in a nuclear power plant and a monitoring system thereof. The monitoring method adopts the technical scheme that a pressure pipe leakage test piece is prepared; the initiation, crack, expansion and through leakage of the pipeline of the nuclear power plant are imitated, a leakage acoustic emission calibration test is performed, and the through crack leakage of the pipeline under high temperature and high pressure environment in nuclear power plant is imitated. The monitoring system adopts a standardized and modularized design which is based on PXI bus and convenient for system maintenance and part replacement. Through adopting the monitoring method and the monitoring system, the data record of leakage event and the database management of monitoring result can be achieved; moreover, the leak position and the dynamic change of leak rate are completely reflected.

The invention belongs to the technical field of automatic control of instruments and meters, and particularly relates to a nuclear power DCS platform test device based on a simulation system and a test method thereof. The device comprises a model server which operates all the system simulation models including chemical and volume control systems; an upper computer of which the function is communication configuration and operates monitoring control frames; and a communication server which realizes data communication between field acquisition controllers and the model server, wherein the port A of the communication server is connected with a lower switch and the port B is connected with an upper switch. The model server is connected with the first port of the upper server, the upper computer is connected with the second port of the model server, and the communication server is connected with the third port of the upper switch. The field acquisition controllers 1-8 are connected with the first to eight ports of the lower switch. The problems that the conventional hardware simulation technology does not perform modeling and cannot realize simulation of the actual physical and technical process control of a power plant can be solved, and the problems that logic testing cannot realize testing according to the technical control requirements can also be solved.

The invention belongs to the technical field of welding materials, and particularly discloses a nickel-based welding wire for main equipment of a nuclear island of an AP1000 nuclear power station. The method is applicable to welding pressure containers of reactors (including welding of driving tube seats, safe seats of connecting tubes and supporting blocks of reactor cores) and welding of steam generators (including build-up welding of tube plates and welding of tubes and the tube plates), and solves problems that welding wires of the kind in the prior art are always imported, cost is high and the like. Basic chemical components of the nickel-based welding wire include, by weight ratio, from 28.0 to 31.5% of Cr, from 7.0 to 11.0% of Fe, from 0.4 to 1.0% of Ti, from 0.25 to 1.10% of Al, from 0.90 to 1.5% of Al+Ti, lower than or equal to 1.0% of Mn, lower than or equal to 0.02% of Nb, lower than 0.04% of C, lower than or equal to 0.15% of Si, lower than 0.005% of P, lower than 0.005% of S, lower than 0.001% of B, lower than 0.02% of Zr, lower than 0.005% of Ca, lower than 0.005% of Mg, lower than 0.02% of Ta, lower than 0.02% of Cu, lower than 0.05% of Co, lower than 0.5% of Mo and the balance Ni. By the aid of the welding wire, microalloying of weld joints can be realized, the weld joints meet standard requirements, and the welding wire can replace imported welding wires.

The invention discloses a method and system for testing a configuration logic design of a DCS (Distributed Control System) of a nuclear power station. The method comprises the following steps of: importing a designed configuration logic diagram to be tested into a selected minimal configuration DCS; communicatively connecting the selected minimal configuration DCS with an engineering simulation process platform; according to the preset test progress, running the configuration logic diagram to be tested on the minimal configuration DCS, and interacting data with the engineering simulation process platform, then recording a test result; and according to the test result, modifying the configuration logic diagram to be tested. According to the method and system disclosed by the invention, the configuration logic diagram is tested by using the minimal configuration DCS and the engineering simulation process platform, the correctness and completeness of the result can be virtually verified without depending on real field equipment, the debugging work period is shortened, and the safety and reliability of the design process is guaranteed.

A nuclear power plant with an improved cooling system using nanoparticles in solid or fluid form is provided. The nanoparticles are delivered in locations such as the cold leg accumulator and high and low pressure pumps of an emergency core cooling system. Motor driven valves and pressurization can aid in the delivery. Methods for providing the nanoparticles are also provided.

The invention relates to a multi-operating mode energy-saving controlled combined type thermostatic and humidistatic air conditioning unit with a thermometal compound box body structure special for nuclear power stations. The air conditioning unit comprises a control system and a box body structure, wherein a primary-efficiency filter, a medium-efficiency filter, a surface cooler, a heater, a humidifier and a blow-off device are sequentially arranged in the box body; the front side of the primary-efficiency filter is provided with a fresh air and return air mixing section, and inlets of fresh air and return air are respectively provided with a fresh air proportioning valve and a return air proportioning valve; the control system adopts a programmable logic controller (PLC) and adopts air conditioning multi-operating mode zone control to realize the energy-saving operation of the unit; the surface cooler adopts a large temperature difference cold water heat exchanging technology and can save the investment and operation expenses of a water system; and the box body adopts a double-layer compound structure and an anti-shock device to meet the special requirements on the nuclear power stations. The invention has the characteristics of high efficiency, energy saving, good tightness, strong anti-shock capacity, safety, reliability, long service life, and the like.

The invention discloses a testing device of a nuclear power station digitized instrument and control system. The testing device relates to the field of automation control and comprises a control device and a signal processing device, wherein the control device comprises an industrial control computer, a collection card installed in the industrial control computer and a control software controlling the testing process; the signal processing device comprises a layplate and a conditioning card; two ends of the layplate are connected with the collection card and the conditioning card through signal wires respectively; and the other end of the conditioning card is connected with a DCS (data communication system) control cabinet to be tested through a signal wire. The testing device adopts the DCS control cabinet or process system as a unit for testing, and can enable all the signals to be simultaneously connected into the testing device for finishing all the test items of the DCS control cabinet or the process system by one time, thus improving testing efficiency and saving human resource. The control method shows the testing state and expected result in real time, the simulation software can simulate the control loop of the process system, and the control accuracy and control effect of the process system state testing can be set according to the needs.

The invention relates to a device and a system for monitoring and displaying normal operating condition of a unit of a nuclear power plant. The nuclear power generating unit has a plurality of operating modes; on a display device, a monitoring module is imported aiming at one operating mode; the monitoring module comprises four units displaying simultaneously, namely a trend tracking unit, a value display unit, a device state unit and an accident condition indication unit. The technical scheme of the invention is that: a reactor operating mode is dominant so as to perform a system monitoring design; according to the change of the normal operating state of the nuclear power generating unit, for a certain working mode, frequent and intersystem monitoring and analysis are needed; parameters and equipment to be monitored are concentrated in a unified monitoring module and displayed through the display device so as to facilitate the monitoring and operation of the total state of the nuclear power generating unit; and the defect of a DCS (digital control system) in the aspect of intuition is overcome, the advantages of a control room of the DCS are fully played, and the safe and stable operation of the nuclear power plant is ensured.

An offshore, floating, moored, nuclear power generating plant and multi-purpose platform is disclosed herein. In a preferred embodiment, the invention consists of a spar or cell spar platform with multi-purpose, all weather topside decks, attached to a submerged “dry tank” that further includes: reactor generator deck(s), power plant main control deck, and central plant deck, that are all integrated within a watertight ballast hull. The invention design further includes cells that are modular for facilitating factory assembly and ultimate construction in a shipyard environment. Reactor vessels are typical naval nuclear reactor having a time tested outstanding safety record. A plurality of reactor generator modules are interconnected and operate independently and collectively and are transmitted by a plurality of high voltage direct current (HVDC) submarine cables through a transformer to the electrical grid. Multipurpose topside decks house vessel commend, crew, and any ancillary and co-generation equipment. The present invention, constructed in a multi-path manufacturing process, provides exceptional economic, environmental, sustainability, security, safety, and operational advantages over the current art of power generation.

The invention relates to a water reducing agent with high efficiency used in polycarboxylate concrete and a synthetic method thereof, which is characterized in that: plyethylene glycol methylacrylic acid single armor esterification (MPEGMA) macromonomer synthesized by (methyl) acrylic and polyethylene glycol monomethyl ether added with catalyst and other (methyl) acrylic or (methyl) allyl sodium monomer are grafted and copolymerizated to synthesize into binary or polybasic copolymer comprising carboxyl, sulfonic group and polyoxyethylene chain side under the effect of initiator and chain transfer agent. The products synthetically prepared by the invention has the advantages of high water reducing ratio, low chlorine and low alkali; the cement mixture is high in fluidity and good in mixability, and can restrain slump loss effectively, when 0.7 to 1.2 percent by weight of cement material is added. The water reducing agent with high efficiency used in polycarboxylate concrete is simple in technology, easy in control of reaction condition and reasonable in cost, does not pollute environment, and can be used as the important component of high performance concrete, marine concrete, nuclear power station concrete and anticorrosion concrete and is widely used in the field of industry, civil building, water conservancy, road traffic and nuclear power engineering.

The invention belongs to a nuclear power station safety system, and in particular relates to a nuclear power station non-active engineering safety system. The system comprises a secondary non-active residual heat removal exchanger, a steam condensation water tank, a non-active reactor cavity water injection system, a non-active high pressure reactor core water supplementing tank and related valves and pipelines. When a nuclear power station is in a design basis accident or a hyper design basis accident, a series of non-active and active safety devices are invested in steps, and a reactor primary circuit and a reactor core are timely, rapidly and effectively cooled, thus a nuclear power station smoothly enters into a safe cold shut-down amplifier state, a reactor disaster consequence is inhibited or alleviated, incident harm is reduced, and the safety of the nuclear power station is improved.

The invention is concerned with an assembled and seal method of box crane beam with ring type orbit using for nuclear power plant, belonging to the establishment construction technology field of nuclear power plant. It relates to rectification of stuff, sealing inner and outside webs and clapboard, sealing upper and down wing edge and the inner and outside webs, sealing stiffened panel, sealing end plate and the rectify after seal. Assemble and seal the web and inner muscle plate to a type of H to reduce the distortion and keep the seal quality, while the temporary backboard controls the final seal distortion when sealing the end plate. It will control the distortion effectively to reach the demand of the first level of nuclear quality and nuclear safety through easy adjustment of circle crane beam after sealed with steady quality.

The invention provides an irradiation-crosslinked low-smoke halogen-free inflaming retarding insulation material for a nuclear power station cable. The insulation material comprises the following components in parts by weight: 100 parts of basic resin (A), 2-5 parts of antioxidant (B), 3-10 parts of irradiation-resisting agent (C), 110-170 parts of fire retardant (D), 2-10 parts of compatilizer (E), 0.1-0.8 part of anticopper agent (F), 0.2-2 parts of lubricant (G) and 2-10 parts of cross-linking sensitizing agent (H), wherein the basic resin (A) comprises a combination of polyethylene-octeneelastomer (POE) ethylene-butyl acrylate (EBA) and linear very low density polyethylene (VLDPE), and the weight ratio of the POE to the EBA to the VLDPE is (10-20):(65-80):(10-20).

Disclosed is a remote control system for a fuel handling system of a nuclear power plant, in which the system includes a remote control console; a programmable logic controller (PLC) unit; a servo motor drive; and a human-machine interface (HMI) unit.