4099 results about "Nuclear power plant" patented technology

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.

An exemplary system method for the operation of a CCP P with flue gas recirculation to reduce NOx emissions and/or to increase the CO2 concentration in the flue gases to facilitate CO2 capture from the flue gases is disclosed. The flue gas recirculation rate (rFRG) is controlled as function of the combustion pressure and/or the hot gas temperature. Operability is enhanced by admixing of oxygen or oxygen enriched air to the gas turbine inlet gases or to the combustion.

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.

A device and method for discovering, identification and monitoring of mechanical flaws in metallic structures is disclosed, based on magneto-graphic/magnetic tomography technique to identify stress-related defects. The technique is specifically optimized for extended, not-accessible underground and underwater metallic structures quality control, emergency alarms as well as timeline planning for structural repairs and maintenance work. Examples of the technique implementation include pipes for oil and gas industry, detection of flaws in rolled products in metallurgical industry, welding quality of heavy duty equipment such as ships, reservoirs, etc. It is especially important for loaded constructions, such as pressured pipes, infrastructure maintenance, nuclear power plant monitoring, bridges, corrosion prevention and environment protection.

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.

The invention discloses an integral low-temperature nuclear heat reactor with the circuit equipment adopting integral arrangement and belonging to the low- and medium-parameter pressurized-water reactor, wherein, a reactor core adopts the mature nuclear power plant fuel component and control rod component; a main heat exchanger is of integral coil type; a voltage stabilizer is a built-in nitrogen partial pressure control voltage stabilizer; coolant circulation is completed by a built-in jet apparatus and the equipment of an external drive circuit; the drive circuit and the equipment and a main circuit auxiliary system are arranged at the circumference of a reactor pressure vessel; a containment vessel consists of a reactor body containment vessel and a reactor top containment vessel; the reactor body containment vessel which is a structure combined by a reactor vault of reinforced concrete structure and a casing of steel structure is connected with a sealed refueling water storage pool through a pipe and a valve. The thermal power of the reactor can be selected between 50MW and 500MW at will and the outlet temperature of the reactor can be selected between 100 DEG C to 200 DEG C according to application, requirement and power.

The invention relates to a safety injection system, which is used for mitigating an accident and protecting the safety of a reactor core of a nuclear power plant and comprises a high-pressure safety injection pump, a low-pressure safety injection pump, a safety injection tank and a refueling water tank for containing safety injection water, wherein the refueling water tank is positioned in a safety shell of the nuclear power plant; the safety injection tank is connected on a direct pressure vessel injection pipeline by a pipeline provided with a control valve; the high-pressure safety injection pump is connected between the refueling water tank and the direct pressure vessel injection pipeline by a pipeline provided with a control valve; and the low-pressure safety injection pump is connected between the refueling water tank and the direct pressure vessel injection pipeline by a pipeline provided with a control valve. The safety injection system has the advantages of reducing cross pipelines of the system, realizing continuous operation, simplifying the system operation, also strengthening mitigating capability of the system for the serious accident and improving safety of the nuclear power plant.

The invention provides a large-scale passive nuclear plant reactor core catcher with bottom water injection and external cooling. The catcher comprises a reactor cavity coating the lower middle part of a reactor pressure vessel, and a reactor cavity concrete soleplate is formed at the bottom of the reactor cavity; a refractory layer is formed on the side surface of the reactor cavity and the bottom of the reactor cavity concrete soleplate; a steel cylinder is sleeved outside the refractory layer; an external cooling passage is formed at the bottom of the steel cylinder, and a cooling passage inlet and a cooling passage outlet are respectively formed at the two outward-extending ends of the external cooling passage; and dozens of nozzles are fixed at the bottom of the steel cylinder, and the upper ends of the nozzles extend into the reactor cavity concrete soleplate while the lower ends of the nozzles extend into the external cooling passage. According to the invention, the dilution and the temperature reduction of a melt are implemented through the melting of concrete by adopting the reactor cavity concrete soleplate as a sacrificial material. A reactor core melt is collected in the refractory layer after the reactor cavity concrete soleplate is molten through, so that the security and the reliability of a nuclear plant are further improved.

The invention relates to the fault tree analysis system for a nuclear power plant with advanced boiling water reactor. The full digital instrument control system uses six different modes to simulate the transmission of the digital signals and the analog signals from the detection units. It is to develop the fault tree for various signal transmission modes to support the nuclear power plant in probabilistic risk assessment (PRA) and meet requirements for simulated signal detection, transmission, logic operation and equipment actuation. Thus, the digital instrument control flow process can fit into PRA model and properly reflect its importance in risk assessment.

The invention discloses a large passive pressurized water reactor nuclear power plant crucible-type reactor core catcher which comprises a reactor cavity concrete bottom plate (4), a crucible cooling system water filling nozzle (6), a crucible cooling system water vapor outlet (7), a crucible component (8), a crucible cooling system cavity (9) and a melt collector (10). The reactor core catcher is organically combined with an interactive voice response (IVR) system, the safety of the nuclear power plant can be further improved, and the reliability of the system is higher due to the crucible design.

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 provides a large-scale passive pressurized water reactor nuclear power plant reactor core catcher with a melt expansion room. The large-scale passive pressurized water reactor nuclear power plant reactor core catcher with the melt expansion room comprises a reactor cavity which coats the middle lower portion of a reactor pressure vessel, a reactor cavity concrete base plate is arranged on the lower portion of the reactor cavity, and a reactor cavity refractory layer is arranged on the lower portion of the reactor cavity concrete base plate. The upper end of a melt release passage is communicated with the reactor cavity refractory layer, and the lower end of the melt release passage is communicated with the melt expansion room. The inner wall of the melt release passage surrounds the refractory layer. An expansion room concrete base plate is arranged on the lower portion of the melt expansion room, an expansion room refractory layer is arranged on the lower portion of the expansion room concrete base plate, and an expansion room outside cooling passage is arranged on the lower portion of the expansion room refractory layer. Two ends of the expansion room outside cooling passage extend outwards and are respectively an outside cooling passage entrance and an outside cooling passage exit. The large-scale passive pressurized water reactor nuclear power plant reactor core catcher with the melt expansion room is used for successively implementing expansion, retention and cooling of the melt when the a pressure container loses efficacy and can strengthen capacity of relieving severe accidents of the large-scale passive pressurized water reactor nuclear power plant.

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.

The invention provides a device combining in-core and out-of-core dwelling of a molten material of a large-scale passive nuclear power plant. The device comprises a concrete sacrificial layer (4), a core catcher chamber (7), a core catcher refractory layer (8), a cooling channel inlet (9), a cooling channel outlet (10) and a core catcher bottom cooling channel (11). The invention provides a set of device organically combining molten material out-of-core cooling with an IVR (interactive voice response) system, when IVR succeeds, in-core dwelling of the molten material can be realized; and after the IVR fails, out-of-core dwelling of the molten material is realized by passive cooling of the core catcher bottom cooling channel, and the capability of mitigating severe accidents of the large-scale passive pressurized water reactor nuclear power plant is further enhanced.

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.

In a start-up method of a single shaft combined cycle power plant, during a time period from when the gas turbine is started up to when rotation speed of the gas turbine reaches a rotation speed allowing self-sustained operation using the combustion gas, auxiliary steam from the start-up boiler is fed to the low-pressure steam turbine via the low-pressure turbine steam supply pipe by controlling valve opening degree of the auxiliary steam flow control valve. The low-pressure steam turbine generates a drive force. Speed-up control is performed in a unified manner for the gas turbine, the high-pressure steam turbine, the low-pressure steam turbine, and the power generator.

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 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 discloses a method for estimating the fatigue life of a pipeline of a nuclear power plant. The method comprises the following steps: S1, measuring the temperature of the outer wall of the pipeline; S2, calculating information on a temperature field of the inner wall of the pipeline; S3, counting the cycle index of an effective load, determining alternating stress, and amending an S-N curve; S4, calculating a fatigue coefficient and amending the fatigue coefficient; and S5, estimating the fatigue life. The method has the beneficial effects of helping a designer master relevant data of the nuclear power plant under a thermal fatigue environment, optimizing an operating standard and an inspection outline according to the estimated fatigue life, developing the design safety margin of the pipeline of the nuclear power plant and providing data support for the regular audit or lifetime extension of the power plant.

The invention is applied in the field of nuclear power control, and provides a method for controlling the starting and stopping of a turbo generator set of a nuclear power plant, a device for controlling the starting and stopping of the turbo generator set of the nuclear power plant, and a digital control system (DCS). The method comprises the following steps of: detecting the running state and running parameters of the turbo generator set, the running state and running parameters of an auxiliary system of the turbo generator set, and fault information in the running of the turbo generator set; displaying the detected information on the same function tracking interface in a centralized way; and directly controlling the starting and stopping of the turbo generator set according to the information displayed on the function tracking interface. The method provided by the embodiment of the invention can make an operator directly and rapidly discover the anomaly of the turbo generator set, rapidly and accurately obtain key parameters and judge reasons, so that the operator can directly, rapidly and accurately control and intervene the starting and stopping of the turbo generator set of the nuclear power plant to ensure the security of the turbo generator set of the nuclear power plant.

The invention relates to an ultrasonic automatic scanning device for a large-aperture pipeline in a nuclear power plant. The ultrasonic automatic scanning device comprises a mechanical scanning part, an electric control part and a pipeline part, wherein the mechanical scanning part comprises a circumferential guide rail part, a circumferential moving module and an axial moving module; the circumferential guide rail part comprises an annular body and a plurality of supporting legs which are mounted on the inner side of the annular body and are capable of moving along the radial direction of the annular body; the circumferential moving module comprises a circumferential guide rail fixed on the annular body, an arc-shaped fixed plate which is detachably mounted on the circumferential guide rail and is capable of moving along the circumferential guide rail, and a circumferential driving motor which is fixed at one end of the arc-shaped fixed plate and is meshed with the circumferential guide rail; the axial moving module comprises an arc-shaped frame detachably mounted on the arc-shaped fixed plate, a probe disk which is parallel to the arc-shaped fixed plate, can slide in the arc-shaped frame and is used for fixing ultrasonic probes, an axial driving motor fixed on the arc-shaped frame, and a transmission mechanism which is connected with the axial driving motor and is used for driving the probe disk. On one hand, the mechanical scanning part is modularized, so that the occupation space can be effectively reduced, and the mechanical scanning part can conveniently move and be mounted in the narrow space of the nuclear power plant; on the other hand, the mechanical scanning part is in a detachable quick connection locking mode, so that the mounting time is shortened.

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.

The invention relates to a simulation system of loading and unloading operation processes in pressurized water reactor (PWR) nuclear power plants. The system comprises an LED (light emitting diode) display screen, a touch screen, an operation control table consisting of a component part consisting of a button, a switch, an operating rod, an indicator lamp and an alarm, an information communication and processing device consisting of a programmable logic controller PLC (programmable logic controller) and a PC graphic workstation, a display device consisting of a scene display and a camera, and an industrial personal computer. The simulation system of the loading and unloading operation processes is designed in strict accordance with functions of real loading and unloading machines, is featured with rich simulated objects, reliable simulation data and strong reality, simulates various common abnormal states or failure states, allows interactive operation, clears faults and has a component showing function; the touch screen has an interface control function; the system simplifies layout of an operation table, expands subsequent functions, effectively trains loading and unloading operators and is widely applied to nuclear power stations and related analogue simulation systems.

The invention discloses a nuclear power halogen-free, low-smoke and flame-retardant cable sheath material with high elongation and long service life and a preparation method thereof, and belongs to the technical field of materials for electric wires and cables. The cable sheath material is characterized by consisting of the following components in parts by weight: 100 parts of a resin base material, 100 to 150 parts of composite flame retardant, 5 to 25 parts of coordinated flame retardant, 3 to 20 parts of a first interface compatilizer, 1.5 to 4 parts of composite antioxidant, and 2 to 5 parts of lubricating agent. The cable sheath material has high mechanical property, particularly elongation at break, good thermo-oxidative aging resistance and long service life; the inherent triangular contradiction of electric, mechanical and flame-retardant properties of the halogen-free, low-smoke and flame-retardant cable sheath material can be effectively solved; meanwhile, the cable sheath material is excellent in processing performance, facilitates extrusion molding, and is scientific and reasonable in formula and proportion and advanced in process.