Abstract

The present invention relates to a fault tree modularization method and apparatus, a storage medium, and an electronic device. The method comprises: obtaining a fault tree; converting the fault tree into a directed graph; initializing the directed graph; performing node traversal on the initialized directed graph to obtain non-modular nodes and leaf nodes; and removing the non-modular nodes and the leaf nodes to obtain modular nodes of the fault tree. According to the present invention, in a fault tree modularization process, the number of traversals for a repeated node is the number of repetitions thereof, and a non-repeated node only needs to be traversed once; compared with existing methods in which three values are recorded for each node, in the present invention, only the father node of each node needs to be recorded, the time complexity of an algorithm is remarkably reduced, node assignment is not needed, the storage space does not need to be occupied, simplification of the fault tree can be effectively implemented, and the objective of improving the fault tree calculation efficiency can also be achieved.

IPC Classes  ?

• G06F 18/20 - Analysing

Abstract

A method for constructing and controlling a turbulence form of a fluid inside a reactor pressure vessel. The method comprises: step S1, obtaining related data of vortexes, a temporal distribution characteristic, a spatial distribution characteristic and a power spectral density of turbulence in an inlet area of each fuel assembly, and a turbulence energy spectrum; step S2, obtaining a water gap change situation of each fuel assembly, and on the basis of the water gap change situation of each fuel assembly, determining whether a three-dimensional flow field model of each fuel assembly needs to be corrected, until a water gap of each fuel assembly does not change; step S3, obtaining a reactor core nuclear power fluctuation amplitude value between fuel assemblies; and step S4, determining whether the reactor core nuclear power fluctuation amplitude value falls within a target range, and if the reactor core nuclear power fluctuation amplitude value does not fall within the target range, changing a three-dimensional structure inside a reactor pressure vessel until the reactor core nuclear power fluctuation amplitude value falls within the target range. The method for constructing and controlling a turbulence form of a fluid inside a reactor pressure vessel can eliminate or ameliorate the nuclear power fluctuation of a reactor core, reduce the wear of grids of the reactor core, and ensure the safety of reactor operation.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G06F 30/28 - Design optimisation, verification or simulation using fluid dynamics, e.g. using Navier-Stokes equations or computational fluid dynamics [CFD]
• G06F 113/08 - Fluids
• G06F 119/14 - Force analysis or force optimisation, e.g. static or dynamic forces

Abstract

A value-taking analysis method for a default value of a signal of a nuclear power station control system, and a terminal device. The value-taking analysis method comprises: acquiring at least one preset default value of a target signal, and failure association data capable of representing actions correspondingly executed by a control system after the target signal fails (S10); on the basis of the failure association data, performing influence degree analysis processing on each preset default value, so as to obtain a discovery difficulty score, an influence degree score and a redundancy degree score (S20); and on the basis of the discovery difficulty score, the influence degree score and the redundancy degree score which correspond to each preset default value, determining a total score corresponding to each preset default value, and determining a settable default value of the target signal on the basis of all total scores (S30). The value-taking analysis method and the terminal device can help a worker to rationally evaluate the rationality of each preset default value of a control signal, and recommend the most rational default preset value to the worker as a recommended value of a default value.

IPC Classes  ?

• G05B 9/03 - Safety arrangements electric with multiple-channel loop, i.e. redundant control systems

Abstract

The present application relates to an overall leakage rate monitoring apparatus for a containment vessel, and an overall leakage rate monitoring method for a containment vessel. The monitoring apparatus comprises: a flow monitoring assembly, a pressure monitoring assembly, a pressure comparison monitoring assembly, an instrument monitoring assembly and a processor, wherein the flow monitoring assembly, the pressure monitoring assembly, the pressure comparison monitoring assembly and the instrument monitoring assembly are each connected to the processor; the flow monitoring assembly is used for monitoring a depressurization flow value inside a containment vessel; the pressure monitoring assembly is used for monitoring a pressure value inside the containment vessel; the pressure comparison monitoring assembly is used for monitoring a pressure comparison value between the inside of the containment vessel and the outside of the containment vessel; the instrument monitoring assembly is used for monitoring the temperature and humidity in the containment vessel; and the processor is used for acquiring the overall leakage rate of the containment vessel on the basis of the depressurization flow value, the pressure value, the pressure comparison value, and the temperature and humidity. The method can solve the defect in an instrument configuration method for an overall leakage rate test for a containment vessel of a nuclear reactor.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting

Abstract

A reactor cavity water injection system and a top-mounted heat preservation module (20) of a reactor pressure vessel thereof. The top-mounted heat preservation module (20) comprises a heat preservation assembly (21) and upper support structures (22); the heat preservation assembly (21) is provided with a heat preservation cavity for a reactor pressure vessel to be placed from the upper end; and the upper support structures (22) are disposed at the upper end of the heat preservation assembly (21) to be hung onto supports on the outer side. The heat preservation module (20) with the upper support structures (22) is prefabricated and assembled outside a nuclear island building and then hoisted as a whole to a reactor pit (10) for adjustment and fixation, and after being hoisted, the upper support structures (22) form a support at the upper end to hang the heat preservation module (20), thereby improving the stability of the heat preservation module (20). The mounting mode is simple and rapid, and the mounting and adjustment duration in the reactor pit (10) is shortened to be within one month, significantly improving the construction efficiency and improving the economy and competitiveness of the unit.

IPC Classes  ?

• G21C 11/08 - Thermal shieldsThermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield
• G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details
• G21C 13/024 - Supporting constructions for pressure vessels or containment vessels

Abstract

The present invention relates to a reactor cavity water injection system and an upper hanging and lower bearing type heat preservation module of a reactor pressure vessel of the reactor cavity water injection system. The upper hanging and lower bearing type heat preservation module comprises a heat preservation assembly, an upper supporting structure and a lower supporting structure, wherein the heat preservation assembly is provided with a heat preservation cavity into which the reactor pressure vessel is placed from an upper end; the upper supporting structure is arranged at the upper end of the heat preservation assembly so as to be hung from an outer support; and the lower supporting structure is arranged on the lower side of the heat preservation assembly and creates support at the bottom of the heat preservation module. An integral heat preservation module with the upper supporting structure and the lower supporting structure is prefabricated outside of a nuclear island, the integral heat preservation module is integrally hoisted to a reactor pit and then adjusted and fixed after being assembled, the upper supporting structure and the lower supporting structure create support at the upper end and the lower end at the same time after being hoisted, such that the stability of the heat preservation module is improved, the mounting is simple and rapid, a mounting and adjustment period in the reactor pit is shortened to be within one month, the construction efficiency is significantly improved, and the economy and competitiveness of a unit are improved.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A cavity injection system, and a bottom-supported heat insulation module (20) for a reactor pressure vessel (30). The bottom-supported heat insulation module (20) comprises a heat insulation assembly (21) and a lower supporting structure (23); the heat insulation assembly (21) is provided with a heat insulation cavity for the reactor pressure vessel (30) to be placed therein via the upper end; and the lower supporting structure (23) is arranged on the lower side of the heat insulation assembly (21) to form support at the bottom of the heat insulation module (20). The integral heat insulation module (20) with the lower supporting structure (23) allows for prefabrication and assembly outside a nuclear island plant and then integral hoisting to a reactor pit (10) for adjustment and fixation, and after hoisting, the lower supporting structure (23) forms support at the lower end.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A zinc injection control method and apparatus, a zinc injection system, a computer device, a computer-readable storage medium, and a computer program product. The method comprises: controlling a zinc injection apparatus to perform zinc injection on a loop of a nuclear power plant at a preset zinc injection flow rate; acquiring a real-time zinc concentration and real-time parameters of a unit of the loop; and on the basis of the real-time zinc concentration, the real-time parameters of the unit of the loop and a standard concentration range, performing control on the zinc injection.

IPC Classes  ?

• G21D 3/08 - Regulation of any parameters in the plant
• G21F 9/00 - Treating radioactively contaminated materialDecontamination arrangements therefor
• G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators

Abstract

A rod swap system and method for a reactor of a nuclear power plant. The rod swap system for a reactor of a nuclear power plant comprises: a power determination unit (1), which is used for determining whether a loop power is stable, and generating a determination result for representing whether the loop power is stable; a power fluctuation prevention unit (2), which is used for monitoring a fluctuation state of the loop power in real time; and a rod swap execution unit (3), which is used for sequentially performing, according to the fluctuation state, rod position swap processing on first and second T rod groups comprised in each temperature rod group after it is determined that the loop power is stable. The rod swap system for a reactor of a nuclear power plant can automatically implement a rod swap operation efficiently, stably and reliably.

IPC Classes  ?

• G21C 7/12 - Means for moving control elements to desired position
• G21C 17/104 - Measuring reactivity
• G21C 17/112 - Measuring temperature

Abstract

The present application relates to a processing method and apparatus for a nuclear facility decommissioning model, a computer device, a storage medium and a computer program product. The method comprises: acquiring three-dimensional models of processed objects; extracting from a pre-established nuclear facility decommissioning information model basic attributes of the processed objects and decommissioning tools, behavioral attributes of the decommissioning tools and engineering attributes of the decommissioning tools; according to action constraint of the decommissioning tools and radiation field dose meshes of the processed objects, performing action simulation on the decommissioning tools; on the basis of a preset interaction mode of the processed objects and the decommissioning tools, simulating action simulation results of the three-dimensional models of the processed objects and the decommissioning tools to obtain target simulation results of the processed objects and the decommissioning tools; and traversing at least two processed objects and decommissioning tools to generate a simulated decommissioning processing solution to the processed objects. The method does not need backstage developers to program various models, improving decommissioning efficiency.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation

Abstract

The present application relates to the technical field of data processing, and in particular to an encoding method and apparatus, a computer device, and a storage medium. The method comprises: acquiring a node index corresponding to a child node to be encoded; determining the node index as a variable value of a Mobius transformation function; and on the basis of a coefficient of the Mobius transformation function and the variable value corresponding to the node index, obtaining an encoding interval corresponding to said child node, wherein the coefficient of the Mobius transformation function is obtained by adjusting a parent interval endpoint value corresponding to the parent node of said child node. The method provided by the present application facilitates construction of an element tree of a device.

IPC Classes  ?

• G06F 8/41 - Compilation

Abstract

The present application discloses a compressed air supply system for a multi-reactor nuclear power plant. The compressed air supply system comprises: at most two compressed air stations (10) connected via a transmission pipe network (20) to N nuclear power units, each compressed air station (10) comprising four parallel-connected primary air compressors (11) and four dryers (12) correspondingly connected to the four primary air compressors (11), the four dryers (12) being arranged in parallel. The invention satisfies the reliability requirements of a downstream user with regard to the compressed air supply system, and reasonably reduces device configuration and cost.

IPC Classes  ?

• G21D 1/00 - Details of nuclear power plant
• F17D 1/04 - Pipe-line systems for gases or vapours for distribution of gas
• F17D 3/01 - Arrangements for supervising or controlling working operations for controlling, signalling, or supervising the conveyance of a product

Abstract

The present invention relates to a drawing table area detection method and apparatus, and a storage medium and an electronic device. The method comprises the following steps: acquiring an original image of a drawing to be subjected to detection; preprocessing the original image of said drawing, so as to obtain a binarized image of said drawing; selecting an initial point from the binarized image of said drawing; on the basis of the initial point, performing inside border detection on the binarized image of said drawing by means of point-by-point pathfinding, so as to obtain an inside border of said drawing; performing table area detection on the basis of the inside border of said drawing, so as to obtain all table areas in said drawing; and performing table extraction on all the table areas in said drawing, so as to obtain all tables in said drawing. By means of the present invention, pixel-level processing is performed on an image, such that a large amount of data labeling work is avoided in a deep learning-based method; model training is performed without consuming a large number of resources, and support from an AI team is not required, such that the cost is low; in addition, the present invention is not interfered with by the content of a drawing, thereby achieving high precision.

IPC Classes  ?

• G06V 30/422 - Technical drawingsGeographical maps
• G06V 30/413 - Classification of content, e.g. text, photographs or tables
• G06V 30/414 - Extracting the geometrical structure, e.g. layout treeBlock segmentation, e.g. bounding boxes for graphics or text

Abstract

A negative pressure centralized monitoring device for a radioactive control area of a nuclear power plant, comprising at least one measurement member (1), a buffer assembly (2), and a pipeline (3). The pipeline (3) comprises a main pipe (31) and at least one branch pipe (32); one end of each measurement member (1) is connected to a control area by means of a branch pipe (32), and the other end of each measurement member (1) is connected to the main pipe (31) by means of the branch pipe (32); the buffer assembly (2) is disposed on the main pipe (31); the measurement member (1) and the buffer assembly (2) are arranged in rooms; one end of the main pipe (31) is connected to the outside; the measurement member (1) is used for measuring the difference between the pressure of the control area and the pressure of the outside; and the buffer assembly (2) is used for filtering out the fluctuation of the external atmosphere. According to the device, the fluctuation of the external atmosphere can be filtered out, and the measurement member (1) is prevented from being affected by the disturbance from the external environment, so that the accuracy of the measurement member (1) is improved, and a ventilation and purification system can accurately adjust the air supply amount and the air intake amount. The stable operation of the ventilation and purification system is guaranteed, and the negative pressure in the control area is kept stable; and additionally, the number of holes punched in the outer wall of the control area can be reduced, thereby reducing the influence of punching holes in the outer wall on the sealing performance of the control area.

IPC Classes  ?

• G01L 13/00 - Devices or apparatus for measuring differences of two or more fluid pressure values

Abstract

The present invention relates to a three-dimensional automatic design method and apparatus for trays and bearers in a gallery, and a medium and a device. The method comprises the following steps: S20, determining design information of the arrangement of trays on a gallery section; S30, acquiring gallery civil work model data, and automatically generating and displaying a tray path on the basis of the gallery civil work model data and the design information of the arrangement of the trays on the gallery section; S40, automatically generating a three-dimensional electrical tray model of a gallery according to the design information of the arrangement of the trays on the gallery section and the tray path; S50, automatically generating a three-dimensional electrical bearer model of the gallery on the basis of the design information of the arrangement of the trays on the gallery section and the three-dimensional electrical tray model; and S60, according to a first preset naming rule, automatically naming and storing the three-dimensional electrical tray model and the three-dimensional electrical bearer model. By implementing the present invention, the three-dimensional design efficiency for trays and bearers in a gallery can be greatly improved, and human errors can be reduced, thereby improving the design quality.

IPC Classes  ?

• G06F 30/13 - Architectural design, e.g. computer-aided architectural design [CAAD] related to design of buildings, bridges, landscapes, production plants or roads
• G06T 17/00 - 3D modelling for computer graphics

Abstract

Disclosed in the present invention is a comprehensive fortification system for cold source safety of a nuclear power plant. The system comprises a first-stage damming unit, a second-stage damming unit, a third-stage damming unit and a fourth-stage damming unit which are sequentially arranged in the direction of a water flow entering the nuclear power plant; the first-stage damming unit comprises a harbor entrance damming net arranged at the intake and at least two harbor entrance buoy monitoring units arranged outside the harbor entrance damming net; the second-stage damming unit comprises a second-stage damming net and a first net-bag damming net which are spaced apart in the water flow direction and are used for carrying out full-section damming on seawater passing through an intake open channel; the third-stage damming unit comprises a second net-bag damming nets; and the fourth-stage damming unit comprises a final net. According to the comprehensive fortification system for cold source safety of a nuclear power plant of the present invention, by means of the sequential arrangement of the four stages of damming units, comprehensive treatment such as monitoring, early warning, damming, cleaning, etc., is carried out on disaster-causing objects such as marine organisms and floating objects carried in an intake water flow, such that the disaster-causing objects are prevented from blocking a nuclear power plant filtering device.

IPC Classes  ?

• E02B 9/04 - Free-flow canals or flumesIntakes

Abstract

The present application relates to a flow-induced vibration test apparatus and method, a computer device, a storage medium, and a product. The apparatus comprises: a flow channel groove, a test simulation body, and a collection device; the flow channel groove comprises a fluid inlet section (120), a fluid mixing section (140), a test section (160), and a fluid outlet section (180) connected in sequence; the fluid mixing section (140) is configured to mix the fluid flowing in from the fluid inlet section (120) to generate a mixed fluid; the mixed fluid flows through the test section (160) and flows out from the fluid outlet section (180); the test simulation body comprises a plurality of non-linear heat transfer pipes (220) and a support assembly (240), and the support assembly (240) is configured to fixedly mount the plurality of non-linear heat transfer pipes (220) in the test section (160); and the collection device is provided on the plurality of non-linear heat transfer pipes (220), and is configured to collect test data when the plurality of non-linear heat transfer pipes (220) undergo flow-induced vibration when the mixed fluid flows into the test section (160).

IPC Classes  ?

• G01M 7/02 - Vibration-testing
• G01M 10/00 - Hydrodynamic testingArrangements in or on ship-testing tanks or water tunnels

Abstract

A reactor vortex suppression and flow distribution device, which is arranged in a reactor pressure vessel (15) and comprises a reactor core lower support plate (3) arranged at the bottom of a reactor core in the pressure vessel (15), and a protruding head (14) connected below the reactor core lower support plate (3). A plurality of through holes allowing a coolant to flow through are formed in the protruding head (14). A first coolant passage (4) is defined between the protruding head (14) and a lower head (5) of the reactor pressure vessel (15). A plurality of spaced disturbing pillars (13) are distributed between the protruding head (14) and the reactor core lower support plate (3), the disturbing pillars (13) defining a plurality of second coolant passages (16) between the protruding head (14) and the reactor core lower support plate (3). The reactor vortex suppression and flow distribution device does not use many parts and has a simple structure; the coolant sequentially flows through the first coolant passage (4), the through holes in the protruding head (14), the second coolant passages (16) and through holes in the reactor core lower support plate (3) and then enters the reactor core, thereby effectively suppressing the generation of vortexes when the coolant flows. In addition, owing to coolant flow redistribution, an evenly-distributed flow can be formed at an inlet of the reactor core.

IPC Classes  ?

• G21C 15/14 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts conducting a hot fluidArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts comprising auxiliary apparatus, e.g. pumps, cameras
• G21C 5/10 - Means for supporting the complete structure
• G21C 3/322 - Means to influence the coolant flow through or around the bundles
• G21C 7/32 - Control of nuclear reaction by varying flow of coolant through the core

Abstract

A method and an apparatus for analyzing the feasibility of adding a repair assembly to a reactor, a device, a medium, and a product. The analysis method comprises: when a fuel rod in a core assembly of a reactor is damaged, replacing the damaged fuel rod with a repair rod, to obtain a repair assembly (S202); increasing power of an undamaged fuel rod adjacent to the repair rod in the repair assembly to a peak rod power of the undamaged fuel rod, placing the repair assembly that has undergone the power increase at a hottest assembly of the reactor core, and limiting an inlet flow at the hottest assembly of the reactor core to a target flow, so as to construct a target working condition (S204); constructing different reactor events in the target working condition, and calculating a departure from nucleate boiling ratio of each event (S206); separately comparing the departure from nucleate boiling ratio of each event with a target limit value, and according to a result of the comparison, determining a feasibility of normal operation of the repair assembly in the reactor (S208).

IPC Classes  ?

• G21C 17/04 - Detecting burst slugs
• G21C 17/10 - Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain

Abstract

A global data control method and apparatus (10) for the network security of an industrial control system of a nuclear power plant, and a central control device and a storage medium. The method comprises: acquiring nuclear power management data of a nuclear power management server cluster (S201), wherein the nuclear power management data comprises network security data of an industrial control system of a nuclear power plant; acquiring operation data of the at least two nuclear power industrial control system clusters by means of a system cluster port-side isolation apparatus (S202), wherein a cluster space isolation apparatus is configured between adjacent nuclear power station system clusters; processing the operation data and the nuclear power management data, so as to obtain global control data of the nuclear power plant (S203); and transmitting, by means of a control-side isolation apparatus, the global control data from a jurisdiction region inside the nuclear power plant to a jurisdiction region outside the nuclear power plant (S204).

IPC Classes  ?

• G05B 19/418 - Total factory control, i.e. centrally controlling a plurality of machines, e.g. direct or distributed numerical control [DNC], flexible manufacturing systems [FMS], integrated manufacturing systems [IMS] or computer integrated manufacturing [CIM]

Abstract

A control method and system for an electric power of a steam turbine generator of a nuclear power plant. The method comprises: S10, collecting a real-time thermal power of a first loop in a nuclear power plant; S20, on the basis of a moving average method, calculating a real-time moving thermal power of the real-time thermal power within a first set time; S30, performing front-end heat generation amount calculation processing on the real-time moving thermal power and the real-time thermal power, so as to obtain a front-end actual heat generation amount; S40, performing a heat generation amount prediction operation according to a preset target average thermal power, the real-time moving thermal power and the front-end actual heat generation amount, so as to obtain a predicted heat generation amount; and S50, performing a conversion and efficiency correction operation on the basis of the predicted heat generation amount, so as to obtain a given electric power value, which is used for controlling an electric power of a steam turbine generator, and then returning to S10.

IPC Classes  ?

• F01D 21/12 - Shutting-down of machines or engines, e.g. in emergencyRegulating, controlling, or safety means not otherwise provided for responsive to temperature
• G01D 3/00 - Measuring arrangements with provision for the special purposes referred to in the subgroups of this group
• G21C 17/00 - MonitoringTesting

Abstract

Disclosed are a multiplexing method and system for a nuclear power plant control system. The multiplexing method for a nuclear power plant control system comprises: acquiring an operating condition of a unit by means of a life monitoring unit of a main control system, the operating condition being determined by means of a non-safety level status and a safety level status of the main control system, as well as a non-safety level status and a safety level status of an auxiliary control system, a mode being switched between the auxiliary control system and the main control system by means of a multi-channel switching system, and the auxiliary control system multiplexing a display terminal of the main control system; on the basis of the operating condition, starting the auxiliary control system by means of the multi-channel switching system, to diversify a human-machine interface panel or expand the monitoring status of a working condition accident panel. The present method can simplify human-machine interface equipment in a control room, reduce project costs, and complete the monitoring and control of normal and accident conditions of a nuclear power plant, thereby effectively ensuring the safety and operation of the nuclear power plant.

IPC Classes  ?

• G21D 3/00 - Control of nuclear power plant

Abstract

Disclosed in the present invention is a multifunctional protection device for a pipeline in a nuclear power plant, the multifunctional protection device comprising an inner pressure bearing layer sleeved on the periphery of the pipeline for positioning and fixing the pipeline. The multifunctional protection device for a pipeline in a nuclear power plant is provided with a buffer layer, an absorption layer and an outer pressure bearing layer connected in sequence from inside to outside with the inner pressure bearing layer as the center, wherein the buffer layer comprises several hollow pipe bundle layers, the several hollow pipe bundle layers being connected around the inner pressure bearing layer to form the buffer layer; the absorption layer comprises several multi-gap portions which are mutually and alternately arranged, several small absorption layer pores being provided on each multi-gap portion; and the outer pressure bearing layer is arranged on the outermost layer of the multifunctional protection device for a pipeline in a nuclear power plant and is used for providing rigid support and preventing leakage in the pipeline . The multifunctional protection device can solve the problems of earthquake resistance, fracture and shock, thermal insulation and quantitative monitoring of medium leakage of an important pipeline in a nuclear power plant, and can replace a damper, an anti-shock limiting member, an anti-jet baffle, a liquid level monitoring device for a pit, etc., to ensure nuclear safety, optimize the design scheme, and reduce the economic cost of dealing with the problems in nuclear power engineering.

IPC Classes  ?

• F16L 57/02 - Protection of pipes or objects of similar shape against external or internal damage or wear against cracking or buckling
• G21D 1/00 - Details of nuclear power plant

Abstract

Disclosed in the present invention is a reactor core measurement instrument guide assembly, comprising: a bottom plate, which is arranged in a cavity defined by means of a top cover of a reactor pressure vessel, wherein an upper side of the bottom plate is provided with instrument guide pipes and instrument guide pipe supporting columns for supporting the instrument guide pipes, and rigid sleeves are provided on a lower side of the bottom plate; and support columns, which are arranged in an upper in-reactor member of a reactor, wherein the support columns are each provided with an accommodating chamber corresponding to the rigid sleeves, and the rigid sleeves can vertically move along an axis relative to the upper in-reactor member, so as to be inserted into or pulled out of the corresponding accommodating chamber. Compared with the prior art, the reactor core measurement instrument guide assembly of the present invention is provided with the rigid sleeves, wherein each of the rigid sleeves has a large diameter and a relatively thick wall, the rigid sleeves have an ideal rigidity, and when a reactor core measurement instrument is inserted downwards, the reactor core measurement instrument can be accurately aligned with the support columns in the upper in-reactor member, thereby ensuring that the reactor core measurement instrument can be successfully inserted into the support columns in the upper in-reactor member.

IPC Classes  ?

• G21C 19/02 - Details of handling arrangements
• G21C 19/19 - Reactor parts specifically adapted to facilitate handling, e.g. to facilitate charging or discharging of fuel elements
• G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
• G21C 17/00 - MonitoringTesting
• G21D 1/02 - Arrangements of auxiliary equipment

Abstract

A water island system of a nuclear power plant. The system comprises a fresh water pretreatment module (11), a softened water treatment module (12), a desalted water treatment module (13), a softened desalted water storage and distribution module (14), a wastewater collection and treatment module (15), a chemical storage and dosing module (16), a chemistry laboratory module (17) and an integrated operation control module (18). The water island system of a nuclear power plant innovatively designs water treatment related systems or sub-items of the nuclear power plant, so as to form a modular design for a plurality of water treatment functional units; fresh water purification treatment, softening and desalting production and distribution, industrial wastewater treatment, domestic sewage treatment, and centralized chemical agent management in the nuclear power plant are realized; a design scheme involving process flow integration, layout design integration and functional partition integration is achieved; and by means of optimizing a system flow and intensively configuring devices, the process flows are smooth, device resources are shared and repeated configurations of the devices are reduced, thereby saving on land, and reducing the project investment.

IPC Classes  ?

• C02F 9/00 - Multistage treatment of water, waste water or sewage

Abstract

A monitoring method and system for an accident working condition of a nuclear power plant set. The method comprises: collecting accident working condition feature parameters, important set safety signals and special safety facility states, which are related to typical accident working conditions (101); analyzing and processing the accident working condition feature parameters, and selecting, by means of screening, abnormal feature parameters that are not within a preset threshold range (102); performing parallel diagnosis on the typical accident working conditions by using a plurality of logic calculation units (103); and displaying diagnosis results of all the typical accident working conditions on an automatic accident working condition diagnosis picture for a set (104). Accident working condition feature parameters, important set safety signals and special safety facility states are monitored in real time, starting accidents or superimposed accidents of a set are automatically diagnosed in parallel, and the accidents are displayed by means of a human-machine interaction interface, thereby assisting an operator in determining and handling accident working conditions of the set.

IPC Classes  ?

• G21D 3/00 - Control of nuclear power plant

Abstract

The present application relates to a mechanical analysis processing method for a nuclear power plant support member, and the method comprises: for a target support member in a nuclear power plant, determining an original pipeline system load of a pipeline system supported by the target support member in a pipeline system coordinate system (step 202); respectively determining a coordinate axis matching each coordinate axis of a global coordinate system of the nuclear power plant from the coordinate axis of an overall local coordinate system of the target support member, and constructing a mechanical analysis model coordinate system close to a direction of the global coordinate system according to a direction of a matched coordinate axis (step 204); constructing a mechanical analysis model of the target support member on the basis of the mechanical analysis model coordinate system (step 206); and converting the original pipeline system load into a target pipeline system load in the mechanical analysis model coordinate system, causing the mechanical analysis model to perform mechanical analysis processing on the basis of the target pipeline system load (step 208).

IPC Classes  ?

• G06F 30/17 - Mechanical parametric or variational design
• G06F 30/20 - Design optimisation, verification or simulation
• G06F 17/00 - Digital computing or data processing equipment or methods, specially adapted for specific functions

Abstract

Supporting piece and driving mechanism pipe base apparatus Disclosed in the present invention are a supporting piece and a driving mechanism pipe base apparatus. The supporting piece comprises a hollow ring body used for being embedded in a driving mechanism pipe base and for a heat sleeve to penetrate through. A first end of the hollow ring body in the axial direction is provided with a connecting part used for being connected to the driving mechanism pipe base and for limiting relative movements in the axial direction and the radial direction. The inner wall surface of a second end of the hollow ring body in the axial direction is a conical surface and is used for supporting the heat sleeve. The supporting piece can ensure that the heat sleeve does not continue to wear the driving mechanism pipe base body. The overall wear life of the driving mechanism pipe base is prolonged, the operation and maintenance requirements of the driving mechanism pipe base are reduced or eliminated, and the unit safety and cost effectiveness are improved. Moreover, the supporting piece has the advantages of being long in service life, and resistant to abrasion and replaceable; the supporting piece can be disassembled and replaced after experiencing excessive wear, and the integrity of the pressure-bearing structure of the pipe base is not affected.

IPC Classes  ?

• G21C 13/032 - Joints between tubes and vessel walls, e.g. taking into account thermal stresses
• G21C 13/06 - Sealing-plugs

Abstract

A control rod guide tube of a nuclear power station, comprising a protective cover (30) and a whole-course guide assembly (40) mounted in the protective cover (30). The whole-course guide assembly (40) comprises at least one whole-course continuous guide element; a whole-course continuous guide channel is provided in the whole-course continuous guide element; and the whole-course continuous guide channel performs whole-course continuous guide on at least one control rod in a single control rod assembly within a stroke range of the guide tube. According to the control rod guide tube of the nuclear power station, discontinuous guide section structures of guide gratings arranged at intervals are changed into whole-course continuous guide sections formed by long duplex pipe structures, such that a control rod stroke protection function is optimized, the risk that the control rod is bent and deformed due to transverse impact of fluid in an upper cavity is reduced, and the integrity of a control rod structure is ensured; meanwhile, the wear resistance of a specific position of the control rod guide tube is improved, such that the service life of the control rod guide tube is greatly prolonged, and later operation costs are reduced.

IPC Classes  ?

• G21C 7/12 - Means for moving control elements to desired position
• G21C 19/10 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements

Abstract

A design software calling method and apparatus, a computer device, and a storage medium. The method comprises: receiving calling information for design software (101); according to the calling information, starting a software starting engine (102); generating an executable file by means of the software starting engine (103); and according to the executable file, calling corresponding design software (104). Unified management and scheduling of design-related analysis software on a local computer or a remote server (including a high-performance computing platform) are realized, thus solving the problem that there are many types of design software and it is difficult for a user to find software; the software starting engine, as a jump, implements a mode for starting design software; in the software starting engine, management and execution of design software are implemented by means of a configuration file mode, so that the calling applicability of the design software in a local computer or a server is improved.

IPC Classes  ?

• G06F 9/445 - Program loading or initiating

Abstract

The present application relates to a method and apparatus for testing the function of a nuclear steam supply system, and a device, a storage medium and a product. The method comprises: according to a first operation state of a system device in a nuclear steam supply system, determining first function state information corresponding to a target nuclear steam supply system function; acquiring a second operation state of a support system, which corresponds to the target nuclear steam supply system function, in the nuclear steam supply system, and determining second function state information of the target nuclear steam supply system function; determining a target system device in which a disaster occurs, determining, by means of probabilistic safety analysis, an associated system device affected by the disaster of the target system device, and according to the target system device and the associated system device, determining third function state information corresponding to the target nuclear steam supply system function; and according to the first function state information, the second function state information and the third function state information, determining a function state of the target nuclear steam supply system function. By means of the method, the detection efficiency and accuracy can be improved.

IPC Classes  ?

• G06F 11/30 - Monitoring
• G21D 3/00 - Control of nuclear power plant
• G21D 3/04 - Safety arrangements

Abstract

Disclosed in the present invention are a neutronics/thermal-hydraulics coupling method and system for a three-dimensional reactor core of a pressurized water reactor. The method comprises: according to section parameter information and by using a nonlinear iterative coarse-mesh finite difference method and a θ method, performing three-dimensional steady-state and transient neutron diffusion calculation; generating a reactor core physical program, and establishing a corresponding physical model according to the material arrangement and geometric dimensions of a reactor core; compiling the reactor core physical program into a dynamic link library, and also storing the physical model in the dynamic link library; calling the dynamic link library, and receiving and obtaining a simulation result output by the reactor core physical program; calling a dynamic link library compiled from a thermal-hydraulic program, and receiving and obtaining a simulation result output by the thermal-hydraulic program; performing exchange transfer on physical parameters between the reactor core physical program and the thermal-hydraulic program; and establishing a mapping relationship between a reactor core physical program mesh and a thermal-hydraulic program mesh according to the material arrangement of the reactor core. By means of the implementation of the present invention, the efficient and accurate external coupling of a reactor core physical program and a thermal-hydraulic program is realized.

IPC Classes  ?

• G06F 30/23 - Design optimisation, verification or simulation using finite element methods [FEM] or finite difference methods [FDM]

Abstract

A blowing-type ventilation cooling tower and a cooling tower arrangement system. The blowing-type ventilation cooling tower (100) comprises a fan chamber (20), an air intake channel (10) formed in an upper end of a side wall of the fan chamber (20) and provided with a downward opening, an air blower (60) arranged in the fan chamber (20), a rain area chamber (30) which is adjacent to and communicates with the fan chamber (20), a heat exchange chamber (40) which is arranged above the rain area chamber (30) and communicates with the rain area chamber (30), an air exhaust channel (50) which is arranged above the heat exchange chamber (40) and communicates with the heat exchange chamber (40), a water spraying device (80) arranged on the top in the heat exchange chamber (40) and used for introducing cooling water, and a water collecting tank (70) which is adjacent to and communicates with the rain area chamber (30), wherein an air outlet (51), away from the heat exchange chamber (40), of the air exhaust channel (50) is staggered from the heat exchange chamber (40). The blowing-type ventilation cooling tower (100) is suitable for an important service water system of a nuclear power plant, and heat exchange of cooling water for cooling is achieved; and the air intake channel and the air exhaust channel are improved, so that flying objects are effectively prevented from entering the tower so as not to damage members in the tower.

IPC Classes  ?

• F28C 1/00 - Direct-contact trickle coolers, e.g. cooling towers
• F28F 25/10 - Component parts of trickle coolers for feeding gas or vapour
• F28F 19/01 - Preventing the formation of deposits or corrosion, e.g. by using filters by using means for separating solid materials from heat-exchange fluids, e.g. filters
• F28F 25/02 - Component parts of trickle coolers for distributing, circulating, or accumulating liquid
• F28F 25/08 - Splashing boards or grids, e.g. for converting liquid sprays into liquid filmsElements or beds for increasing the area of the contact surface
• F28F 25/06 - Spray nozzles or spray pipes
• F28F 25/04 - Distributing or accumulator troughs
• G21D 1/02 - Arrangements of auxiliary equipment

Abstract

A steam stable supply system and a steam stable supply method. The steam stable supply system comprises a solar heat collection tower (10), a temperature and pressure reducer (30), steam heat storage tanks (20), a first steam pipeline (41) and a first heat metering device (51), wherein the first heat metering device (51) is arranged at a steam outlet end of the solar heat collection tower (10), and measures the steam output quantity of the solar heat collection tower (10); the first steam pipeline (41) is connected between the steam outlet end of the solar heat collection tower (10) and the temperature and pressure reducer (30), and is used to convey steam to the temperature and pressure reducer (30); the steam heat storage tanks (20) are connected to the first steam pipeline (41) in parallel and connected between the steam outlet end of the solar heat collection tower (10) and the temperature and pressure reducer (30), and are used to store steam heat energy and convey, to the temperature and pressure reducer (30), steam generated by flash evaporation of water stored in the steam heat storage tanks (20); and the temperature and pressure reducer (30) is connected to a heat user side (100), and same reduces the temperature and pressure of the steam and then supplies the steam to the heat user side (100). The requirements of highly efficient heat exchange, long-term heat storage, and sustainable and stable supply of steam are satisfied, and the requirements for industrial production are satisfied.

IPC Classes  ?

• F22B 33/18 - Combinations of steam boilers with other apparatus
• F22D 11/02 - Arrangements of feed-water pumps
• F24D 1/08 - Feed-line arrangements, e.g. providing for one-pipe system

Abstract

The present invention relates to an automatic low-pressure full-speed cooling method and system for a nuclear power plant unit. The method comprises: acquiring a unit operation parameter of a nuclear power unit and a state parameter of an important safety device, wherein the unit operation parameter comprises a primary-loop water loading amount and a residual heat removal state function parameter, and the state parameter of the important safety device comprises a state parameter of medium-pressure safety injection and a state parameter of a steam generator; performing logic calculation according to the unit operation parameter and the state parameter, and outputting a logic processing result; triggering an automatic low-pressure full-speed cooling signal according to the logic processing result; and executing automatic low-pressure full-speed cooling according to the automatic low-pressure full-speed cooling signal. By means of the present invention, on the basis of a logic processing result, the accident condition of a large break coincident with a medium-pressure safety injection failure in a primary loop can be monitored, and when the accident condition is detected, an action of low-pressure full-speed cooling can be automatically executed, thereby quickly alleviating the accident condition, preventing the execution risk caused by human factors, and improving the safety margin of a unit under the accident condition.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21D 3/06 - Safety arrangements responsive to faults within the plant

Abstract

A gas-liquid linkage actuating mechanism for a main-steam isolation valve of a nuclear power plant, suitable for three columns of independent safety-level power supply and instrument control systems, and comprising: a gas-liquid linkage piston cylinder (1) connected to a main-steam isolation valve (100), a pump-side quick-closing circuit (2) and a non-pump-side quick-closing circuit (3) that are respectively connected to a hydraulic cylinder (12) of the gas-liquid linkage piston cylinder (1), and a liquid supply apparatus (4). The pump-side quick-closing circuit (2) comprises a first main pipeline (21) having one end connected to the liquid supply apparatus (4) and the other end connected to the hydraulic cylinder (12) of the gas-liquid linkage piston cylinder (1); the first main pipeline (21) is provided with a first quick-closing oil drain valve (22); the pump-side quick-closing circuit (2) further comprises a first control assembly controllably connected to the first quick-closing oil drain valve (22); and the first control assembly comprises three quick-closing solenoid pilot valves connected to each other, and each quick-closing solenoid pilot valve is connected to each column of safety-level power supply and instrument control system. The mechanism solves the problem of adapting to three series of independent safety-level power supply and instrumentation control systems of a nuclear power plant, and also has the inherent safety.

IPC Classes  ?

• F16K 31/122 - Operating meansReleasing devices actuated by fluid the fluid acting on a piston
• F15B 11/08 - Servomotor systems without provision for follow-up action with only one servomotor

Abstract

An emergency residual heat removal and water replenishing system for a nuclear power plant. The system comprises three columns of mutually independent cooling series (10), an ultra-large-sized modular cooling water tank (40), a set of control and alarm systems, and an auxiliary monitoring instrument, wherein the water capacity of the ultra-large-sized modular cooling water tank (40) at least can meet the removal of heat of a loop within 72 hours after an accident occurs, and the cooling water tank (40) is connected to an emergency water supply system and a spent fuel pit, so as to replenish the emergency water supply system and the spent fuel pit with water when needed; the three columns of cooling series (10) share the ultra-large-sized modular cooling water tank (40) as a final heat sink of condensers (14) of the three columns of cooling series, thereby realizing, when the emergency water supply system is not available, the passive residual heat removal at secondary sides of steam generators (90); and the bottom of the ultra-large-sized modular cooling water tank (40) is provided with sinking compartments, and the condensers (14) of the three columns of cooling series (10) are respectively located in different sinking compartments. The system can completely meet the requirement for removing, within 72 hours, heat released by nuclear fuel, thereby improving the capability and safety of the nuclear power plant in coping with an accident condition.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

Disclosed in the present invention is a radiation shielding and heat insulation device, comprising an annular top plate, which is fitted on the top of a container; a shielding unit, which is connected below the annular top plate, arranged on the periphery of the container in a surrounding manner and used for shielding neutrons and γ rays; and a heat insulation layer, which is arranged below the top plate and fixedly attached to an inner side of the shielding unit so as to coat the periphery of the container. The radiation shielding and heat insulation device in the present invention is used for covering an upper portion of a spent fuel transfer container or transport container, has the functions of radiation protection and high-temperature heat insulation, and solves the technical problems of a high radiation dose/radioactive substance contamination, a high operation temperature, etc. that are present during a spent fuel loading operation.

IPC Classes  ?

• G21F 5/015 - Transportable or portable shielded containers for storing radioactive sources, e.g. source carriers for irradiation unitsRadioisotope containers
• G21F 5/06 - Details of, or accessories to, the containers

Abstract

A support structure of a voltage stabilizer of a nuclear power plant. The support structure is provided in the middle of the voltage stabilizer (1), and comprises lugs (10) fixedly connected to the voltage stabilizer (1) and a support base (20) for fixedly connecting the voltage stabilizer (1) and a supporting floor, and the support base (20) is detachably connected to the lugs (10); the support base (20) and the voltage stabilizer (1) are provided in a gap manner. A single-layer support structure is provided in the middle of the voltage stabilizer (1), the support structure is simplified, the inherent frequency of a device is improved, and a response of the device under an earthquake working condition is reduced. Moreover, a certain gap is formed between the support base (20) and the voltage stabilizer (1), so that radial displacement generated due to thermal expansion of the cylinder body of the voltage stabilizer (1) can be absorbed, and rigid constraint of the support base (20) for the cylinder body under the thermal expansion working condition is prevented.

IPC Classes  ?

• G21C 1/09 - Pressure regulating arrangements, i.e. pressurisers
• G21C 13/024 - Supporting constructions for pressure vessels or containment vessels

Abstract

The present invention relates to a straight pipeline inner wall surface temperature measurement and transient identification method and a computer terminal. The method comprises: acquiring measurement temperature data of the outer wall surface of a pipeline, and on the basis of the measurement temperature data of the outer wall surface, using a preset method to carry out thermal conduction inversion calculation to obtain temperature data of the inner wall surface of the pipeline. According to the present invention, the temperature of the outer wall surface can be accurately obtained in real time without damaging a loop pipeline; the temperature of the inner wall surface is calculated on the basis of a thermal conduction inversion analysis method, and reliable input data is provided for thermal stress calculation for fatigue evaluation; repeated iteration is not needed, and the problem of non-convergence does not exist; the calculation time is short, and the calculation amount is small; and temperature information of the inner wall surface of the pipeline can be accurately and quickly identified.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation
• G01K 1/14 - SupportsFastening devicesArrangements for mounting thermometers in particular locations
• G06F 113/14 - Pipes
• G06F 119/08 - Thermal analysis or thermal optimisation

Abstract

A nuclear power plant three-dimensional wayfinding navigation method and system, the nuclear power plant three-dimensional wayfinding navigation method comprising: obtaining a three-dimensional layout model of a nuclear power plant (S10); constructing a three-dimensional virtual scene of the nuclear power plant, and in the three-dimensional virtual scene, dividing the plant according to layers and arranging a relay point in each layer so as to bake a navigation mesh layer by layer (S20); in the three-dimensional virtual scene, positioning a target item inputted by a user and receiving a start point and an end point set by the user, and on the basis of the layered navigation mesh, performing a wayfinding calculation in the three-dimensional virtual scene and performing a dynamic navigation demonstration of an optimal path according to a wayfinding calculation result (S30).

IPC Classes  ?

• G01C 21/20 - Instruments for performing navigational calculations

Abstract

A molten salt reactor capable of realizing online refueling, and a refueling method. The molten salt reactor capable of realizing online refueling comprises a protection vessel (10), and a reactor vessel (20), a reactor core, a supporting mechanism (30) and a hoisting mechanism (40), which are arranged in the protection vessel (10), wherein the supporting mechanism (30) comprises a plurality of supporting rails (31) which are arranged at intervals in parallel and stretch over the reactor vessel; the reactor core comprises a plurality of fuel assemblies (50) arranged in the lengthwise direction of the supporting rails (31); each of the fuel assemblies (50) comprises a fuel grating (51) and TRISO fuel balls (52) accommodated in the fuel grating along the height of the fuel grating (51); the tops of the fuel gratings (51) are suspended on the supporting rails (31); and the hoisting mechanism (40) is located above the supporting mechanism (30). According to the molten salt reactor capable of realizing online refueling, the TRISO fuel balls (52) are loaded into the vertical fuel grating (51) to form a single fuel assembly (50), and the fuel assembly fits the supporting mechanism (30) and the hoisting mechanism (40) above the fuel assembly (50) to realize the arrangement and movement of the fuel assembly (50) in the reactor vessel (20), thereby realizing online refueling of the molten salt reactor, and fine axial enrichment adjustment, modular construction, and power extension of the fuel assembly (50).

IPC Classes  ?

• G21C 1/03 - Fast fission reactors, i.e. reactors not using a moderator cooled by a coolant not essentially pressurised, e.g. pool-type reactors
• G21C 3/02 - Fuel elements
• G21C 5/06 - Means for locating or supporting fuel elements

Abstract

The present invention relates to a draw-in and cast device for a nuclear power circulating water prefiltration system, comprising cast cables connected onto net bags, net draw-in cables connected onto the net bags, tensioning cables connected onto the net bags, a draw-in and cast platform, first draw-in and cast units arranged on the draw-in and cast platform, second draw-in and cast units arranged on the draw-in and cast platform, and third draw-in and cast units arranged on the draw-in and cast platform. The first draw-in and cast units are connected to the net draw-in cables so as to drive the net draw-in cables to be lengthened or shortened to implement net cast or net draw-in; the second draw-in and cast units are connected to the cast cables so as to drive the cast cables to be lengthened or shortened to implement net draw-in or net cast; the third draw-in and cast units are connected to the tensioning cables so as to drive the tensioning cables to be lengthened or shortened to assist net cast by the cast cables. The draw-in and cast device for a nuclear power circulating water prefiltration system achieves full-section interception for the filtration system, so that the filtration system is effectively protected, and the efficiency of replacing the filtration system can be improved.

IPC Classes  ?

• B01D 29/23 - Supported filter elements arranged for outward flow filtration
• B01D 29/52 - Filters with filtering elements stationary during filtration, e.g. pressure or suction filters, not covered by groups Filtering elements therefor with multiple filtering elements, characterised by their mutual disposition in parallel connection
• B01D 29/96 - Filters with filtering elements stationary during filtration, e.g. pressure or suction filters, not covered by groups Filtering elements therefor in which the filtering elements are moved between filtering operationsParticular measures for removing or replacing the filtering elementsTransport systems for filters

Abstract

A flue gas treatment system for hazardous waste incineration, and a flue gas treatment method for hazardous waste incineration. The flue gas treatment system for hazardous waste incineration comprises a rotary kiln (10), a secondary combustion chamber (20), a waste heat boiler (30), a quench tower (40), a dry deacidification tower (50), a dust collector (60), a washing tower (70), a wet deacidification tower (80) and a shunt flue; the rotary kiln (10), the secondary combustion chamber (20), the waste heat boiler (30), the quench tower (40), the dry deacidification tower (50), the dust collector (60), the washing tower (70) and the wet deacidification tower (80) are sequentially connected in the traveling direction of the flue gas; and the shunt flue is connected between an outlet of the dust collector (60) and the rotary kiln (10). The flue gas treatment system treats the flue gas generated by the incineration of the rotary kiln (10), so as to enable the flue gas to meet the requirements of low-nitrogen emission, and solve the problem of heat recovery of the flue gas after the incineration of the rotary kiln (10); and part of the flue gas is guided, by means of the shunt flue, into the rotary kiln (10) for recycling in the subsequent treatment, increasing the air inlet temperature of the rotary kiln (10), and reducing the oxygen concentration of the rotary kiln (10), thereby reducing the emission concentration of nitrogen oxides.

IPC Classes  ?

• F23G 5/14 - Methods or apparatus, e.g. incinerators, specially adapted for combustion of waste or low-grade fuels including supplementary heating including secondary combustion
• F23G 5/20 - Methods or apparatus, e.g. incinerators, specially adapted for combustion of waste or low-grade fuels with combustion in rotating or oscillating drums
• F23G 5/44 - Methods or apparatus, e.g. incinerators, specially adapted for combustion of waste or low-grade fuels DetailsAccessories
• F23G 5/46 - Recuperation of heat
• F23J 15/02 - Arrangements of devices for treating smoke or fumes of purifiers, e.g. for removing noxious material

Abstract

A spent fuel storage unit (200) and a stacked spent fuel storage device, the spent fuel storage unit (20) comprising a storage module (21), and a plurality of corner posts embedded into a plurality of edges of the storage module (21) respectively. An accommodating cavity (210) is provided inside the storage module (21); the storage module (21) is provided with a front surface and a back surface that face each other. The plurality of corner posts comprise a first corner post (22) corresponding to the front surface of the storage module (21) and a second corner post (23) corresponding to the back surface of the storage module (21). The first corner post (22) is provided with an air intake hole (24) in communication with the accommodating cavity (210), and the back surface of the storage module (21) is provided with an air outlet hole (26) in communication with the accommodating cavity (210). The air intake hole (24), the accommodating cavity (210), and the air outlet hole (26) are sequentially connected to form a heat dissipation channel that allows air to pass through and carries away decay heat from a fuel assembly. By means of the improved design of the spent fuel storage unit (20), stacking arrangement with the spent fuel storage unit (20) as the basic unit is made possible, and meanwhile, each storage unit (20) is able to implement distributed and independent air intake and centralized exhaust, thus enabling multi-level stacked assemblies and dense storage of spent fuel having a high thermal load.

IPC Classes  ?

• G21C 19/06 - Means for supporting or storing fuel elements or control elements
• G21C 19/07 - Storage racksStorage pools

Abstract

The present invention belongs to the field of nuclear power, and relates to a power supply apparatus and method based on a high-speed flywheel, and a related device. The power supply apparatus supplies power to a control rod driving mechanism, and comprises a high-speed flywheel, a permanent magnet synchronous electric motor, a bidirectional converter, a rectifier and an inverter output unit, wherein the permanent magnet synchronous electric motor is respectively electrically connected to the high-speed flywheel and the bidirectional converter; every two of the bidirectional converter, the rectifier and the inverter output unit are electrically connected; and the rectifier is connected to a plant power, and the inverter output unit is connected to the control rod driving mechanism. The power supply method comprises identifying the current power supply state in real time; when the current power supply state indicates normal power supply, transmitting a plant power to a control rod driving mechanism and a high-speed flywheel by means of a rectifier and an inverter output unit; and when the current power supply state indicates abnormal power supply, driving, by means of the high-speed flywheel, a permanent magnet synchronous electric motor to rotate, and transmitting, to the control rod driving mechanism, a current generated by means of the rotation of the permanent magnet synchronous electric motor. The power supply apparatus in the present application has a high power supply efficiency, a small occupied area and is convenient to maintain.

IPC Classes  ?

• H02J 3/30 - Arrangements for balancing the load in a network by storage of energy using dynamo-electric machines coupled to flywheels
• H02J 9/08 - Circuit arrangements for emergency or stand-by power supply, e.g. for emergency lighting in which the distribution system is disconnected from the normal source and connected to a standby source with automatic change-over requiring starting of a prime-mover
• H02M 7/66 - Conversion of AC power input into DC power outputConversion of DC power input into AC power output with possibility of reversal

Abstract

A nuclear power plant intelligent technical specification method and system, the method comprising: obtaining a predetermined operating signal and a device state of a nuclear power plant by means of a nuclear power plant KNS system (S101); determining an operating condition of the nuclear power plant according to the device state (S102); determining whether to trigger a predetermined event according to the operating signal, the operating condition, and a preset threshold (S103); upon determining to trigger a predetermined event, activating the predetermined event, and triggering processing measures and supervision requirements corresponding to the event, and simultaneously starting a timer to remind an operator to complete the processing measures of specification requirements within a specified time (S104). The present nuclear power plant intelligent technical specification method and system may intuitively reflect safety indicators of a nuclear power plant, helping an operator to quickly handle an event, so that the safety of the nuclear power plant is improved, the economy and the availability of the unit are improved, and the market economy value is high.

IPC Classes  ?

• G21D 3/00 - Control of nuclear power plant
• G21D 3/06 - Safety arrangements responsive to faults within the plant

Abstract

A human reliability evaluation method and system, and a computer device and a storage medium, which belong to the technical field of system engineering reliability analysis. The method comprises: acquiring a task sequence that comprises a diagnosis task sequence and an action task sequence (S201); performing correction processing on the task sequence to acquire a diagnosis corrected shaping factor corresponding to the diagnosis task sequence and an action corrected shaping factor corresponding to the action task sequence (S202); and evaluating a human error probability according to a corrected performance shaping factor (S203). In the method, by means of analyzing a task sequence and acquiring a corrected performance shaping factor by means of correction processing, the problems in the prior art of qualitative analysis of a performance shaping factor being missing and unclear, qualitative analysis and quantitative analysis not being tightly combined, and credible digital human-machine interface basic data being absent are solved, and the problem of an evaluation process involved in an existing technical solution being ambiguous is effectively overcome, thereby effectively improving the credibility of human factor safety evaluation for nuclear power station design.

IPC Classes  ?

• G06Q 10/06 - Resources, workflows, human or project managementEnterprise or organisation planningEnterprise or organisation modelling

Abstract

A hydrogen concentration measuring device resistant to high temperatures, high pressure, high humidity and high radiation, and a hydrogen measuring probe (20). The hydrogen measuring probe (20) comprises a probe housing, a hydrogen measuring element (24) and a fixing assembly (29), wherein the probe housing is provided with a hydrogen measuring inlet covered with a filter screen (213), and the fixing assembly (29) is used for fixing the hydrogen measuring element (24) in the probe housing; and the hydrogen measuring element (24) is used for measuring the concentration of hydrogen by means of a catalytic electrochemical method and comprises a measuring element housing (241) internally provided with an insulating layer (240), a filtering permeable membrane (242), a measuring electrode (243), a counter electrode (244), a reference electrode (245), an electrolyte layer (246), an oxygen storage layer (248) and a hydrogen measuring element lead (249), the oxygen storage layer (248) being made of metal oxygen storage, oxygen release and oxygen production materials and capable of continuously releasing oxygen. In the hydrogen concentration measuring device, the hydrogen measuring probe (20) is used to measure hydrogen partial pressure. The device is applicable to hydrogen concentration measurement in high-temperature, high-pressure, high-humidity and high-radiation environments of a nuclear power plant, and is also applicable to hydrogen concentration measurement in other severe environments.

IPC Classes  ?

• G21C 17/003 - Remote inspection of vessels, e.g. pressure vessels
• G01N 27/26 - Investigating or analysing materials by the use of electric, electrochemical, or magnetic means by investigating electrochemical variablesInvestigating or analysing materials by the use of electric, electrochemical, or magnetic means by using electrolysis or electrophoresis
• G01N 27/40 - Semi-permeable membranes or partitions

Abstract

Disclosed is a transportation container for a nuclear power station spent fuel storage tank. The transportation container comprises: a container body, which comprises an inner cylinder body and an outer cylinder body sheathed outside the inner cylinder body, with a guide rail being arranged on the inner cylinder body; a top cover, which is mounted at the top of the container body by means of an upper flange; and a penetrating cover plate, which is detachably mounted at the bottom of the container body by means of a lower flange. An upper buffer base plate is arranged between the top cover and a spent fuel storage tank located in the inner cylinder body, and a lower buffer base plate is arranged between the penetrating cover plate and the spent fuel storage tank located in the inner cylinder body. Compared with the prior art, the transportation container for a nuclear power station spent fuel storage tank in the present invention is suitable for bearing horizontal and vertical spent fuel storage tanks, and the spent fuel storage tanks can be directly stored in the transportation container and transported to a post-treatment plant or a geological treatment facility for a long distance, so as to achieve recycling and retreatment of spent fuel assemblies, such that the technical problem whereby spent fuel storage tanks are not able to be directly transported for a long distance domestically and abroad to date is effectively solved.

IPC Classes  ?

• G21F 5/06 - Details of, or accessories to, the containers
• G21F 5/12 - Closures for containersSealing arrangements

Abstract

Disclosed is a composite electrode for a cadmium zinc telluride radiation detector. The composite electrode comprises: a NiTe-based thin-film electrode grown on the surface of a cadmium zinc telluride crystal material, a Ni thin-film electrode grown on the surface of the NiTe-based thin-film electrode, and an Au thin-film electrode grown on the surface of the Ni thin-film electrode. Compared with the prior art, the composite electrode for a cadmium zinc telluride radiation detector of the present invention can form better ohmic contact on the surface of a cadmium zinc telluride crystal material and have a lower contact resistance, which significantly improves the charge collection performance of a cadmium zinc telluride radiation detector. In addition, also disclosed is a method for preparing a composite electrode for a cadmium zinc telluride radiation detector. Compared with the prior art, by means of the method for preparing a composite electrode for a cadmium zinc telluride radiation detector of the present invention, a NiTe-based thin film electrode is prepared by using a magnetron sputtering technique, such that the crystallization quality is good, the adhesive force is strong, the conductivity is high, the speed is high, the quality is stable, and the cost of batch growth by means of the magnetron sputtering technique is low.

IPC Classes  ?

• H01L 31/0224 - Electrodes

Abstract

An ultra-real-time calculation method for axial power deviation, an electronic device, and a storage medium, the method comprising: setting a boundary condition (S101); determining interface data on the basis of the boundary condition (S102); initializing a simulated reactor core physical model based on the interface data and in combination with real-time data (S103); using the interface data in the simulated reactor core physical model to performing ultra-real-time calculation (S104); and outputting and displaying the result of the ultra-real-time calculation (S105). Using the described method, a trend for a next half-hour may be predicted, and a preview and verification platform that provides an axial power adjustment response scheme in the event of an accident for nuclear power plant operators is able to predict dynamic characteristics of the reactivity of a power plant and the impact on neutron flux distribution during accident conditions relatively accurately, simulating in advance the course of development of various accident conditions, and predicting various risks for a later period, thus allowing the operators more sufficient time to reduce or even eliminate risk.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation

Abstract

A passive hydrogen elimination device for a nuclear power plant, comprising a carrier (10) and silver zeolite particles (20) arranged on the carrier (10), wherein the carrier (10) is provided with a chimney-shaped flow channel, and the silver zeolite particles (20) are arranged in the chimney-shaped flow channel or attached to the outer surface of the carrier (10). The invention also relates to a passive hydrogen elimination system for a nuclear power plant using the passive hydrogen elimination device for a nuclear power plant. The passive hydrogen elimination device and system for a nuclear power plant use a method including silver zeolite catalysis and passive principle hydrogen elimination, which realizes reliable hydrogen elimination after an accident, especially a serious accident, solves the problem of the risks of hydrogen burning and hydrogen explosions in a containment of a large space of a nuclear power plant after an accident, and maintains the integrity of the containment; and further realizes reliable hydrogen elimination in a small space and a process system pipeline, and provides reliable methods for hydrogen recombination in a small space, hydrogen elimination in a container, hydrogen elimination in a hydrogen-containing medium during a process, hydrogen elimination in a laboratory, and hydrogen elimination in a regeneration test device of a hydrogen recombiner of a nuclear power plant.

IPC Classes  ?

• G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
• B01J 29/068 - Noble metals
• G21F 9/02 - Treating gases
• B01D 53/86 - Catalytic processes

Abstract

A storage tank cleaning and decontamination system for collecting and processing a radionuclide on the inner surface of a storage tank (1). The system comprises: a spray assembly (100) temporarily introduced into the storage tank (1), for controllably spraying an original detergent and/or the detergent resultant after collection and filtration onto the inner surface of the storage tank (1), and flushing the radionuclide from the inner surface of the storage tank (1) into a filtration and collection assembly (200); the filtration and collection assembly (200) arranged outside the storage tank (1) and directly facing a discharge port (11) of the storage tank (1), for filtering out the radionuclide carried in the detergent, so as to store the original detergent and/or the detergent resultant after collection and filtration; a pump (300) arranged outside the storage tank (1) and connected to the spray assembly (100) and the filtration and collection assembly (200) via a pipeline (500), for pressurizing the original detergent and/or the detergent resultant after collection and filtration and transporting the same to the spray assembly (100). Additionally, the present application further includes a storage tank cleaning and decontamination method. The system transfers the radionuclide or other impurities from the inner wall of the storage tank (1) into a filter element for centralized processing, so as to reduce the difficulty of processing.

IPC Classes  ?

• B08B 9/08 - Cleaning of containers, e.g. tanks
• B08B 9/093 - Cleaning of containers, e.g. tanks by the force of jets or sprays
• B08B 13/00 - Accessories or details of general applicability for machines or apparatus for cleaning
• B01D 29/35 - Self-supporting filtering elements arranged for outward flow filtration

Abstract

A three-dimensional model application method and system for the nuclear power engineering modular construction. The three-dimensional model application method comprises: establishing several nuclear power plant modular three-dimensional models; extracting each technical document from each of the nuclear power plant modular three-dimensional models, and generating purchasing, construction, and debugging information of each nuclear power plant modular three-dimensional model according to the technical document; performing lightweight construction on the several nuclear power plant modular three-dimensional models and generating several lightweight three-dimensional module models; and establishing the correspondence between the several lightweight three-dimensional module models and the purchasing, construction, and debugging information. By using the three-dimensional module models as the core, the method realizes the nuclear power engineering project service process monitoring and management, and refined and dynamic management of the project progress and material management.

IPC Classes  ?

• G06T 17/00 - 3D modelling for computer graphics
• G06Q 50/06 - Energy or water supply

Abstract

A measurement method and a measurement system for nuclear power plant radiation dose distribution: by means of setting up a certain number of video cameras in combination with an existing wireless communication system (30), radiation doses can be measured in real time by means of a mobile dose measuring apparatus (20) carried by all of the staff members in a nuclear power plant, and a radiation dose distribution map in the areas of the nuclear power plant that require dose measurement can be accurately drawn. Thus, in the case of unexpected excess radiation doses, early warning information can be promptly issued, prompting staff members to stay away from related areas that endanger occupational health and safety.

IPC Classes  ?

• G01T 1/02 - Dosimeters

Abstract

A loosening part capturing device for a steam generator (10) of a pressurized water reactor nuclear power plant, and a vertical type steam generator (10) of the pressurized water reactor nuclear power plant comprising same. The loosening part capturing device is provided on a top plate (51) of a sludge collector (50); the top plate (51) is provided with a plurality of steam and water separator rising cylinders (41); capturing enclosing plates (80) fixedly connected to the top plate (51) are provided between every adjacent steam and water separator rising cylinders (41) on the periphery; two ends of each capturing enclosing plate (80) are respectively fixedly connected to outer surfaces of the water separator rising cylinders (41); an end, away from the top plate (51), of each capturing enclosing plates (80) is provided with a folding plate (81) extending towards the center of the top plate.

IPC Classes  ?

• F22B 1/16 - Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being hot liquid or hot vapour, e.g. waste liquid, waste vapour
• F22B 37/22 - DrumsHeadersAccessories therefor
• F22B 37/26 - Steam-separating arrangements

Abstract

The present invention provides a device gate for a nuclear power plant, comprising a shell ring fixedly connected to a penetrating member of a safety housing. The device gate for the nuclear power plant further comprises: a gate end enclosure provided corresponding to an end surface of the shell ring and used for being sealedly connected to the shell ring; a hoisting lifting assembly provided above the gate end enclosure, movably connected to the gate end enclosure, and used for performing hoisting and resetting on the gate end enclosure; a high-position suspending assembly provided above the gate end enclosure and used for being detachably connected to the gate end enclosure when the gate end enclosure is hoisted to a specified position; and an anti-seismic limiting assembly partially provided on the safety housing, movably connected to the gate end enclosure, and used for performing limiting and anti-seismic processing on the gate end enclosure when the gate end enclosure is suspended on the high-position suspending assembly. The present invention ensures reliability and anti-seismic requirements of the device gate for the nuclear power plant.

IPC Classes  ?

• F16K 3/02 - Gate valves or sliding valves, i.e. cut-off apparatus with closing members having a sliding movement along the seat for opening and closing with flat sealing facesPackings therefor
• F16K 3/316 - Guiding of the slide
• B66F 7/06 - Lifting frames, e.g. for lifting vehiclesPlatform lifts with platforms supported by levers for vertical movement
• G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
• B66D 1/04 - Driving gear manually operated

Abstract

The present invention provides a spent fuel assembly uncovering and recycling system and method, and a spent fuel storage tank (1) suitable for storing a spent fuel assembly. The spent fuel assembly uncovering and recycling system comprises: a helium leakage and cladding damage detection assembly (100) connected to the spent fuel storage tank (1) and used for controllably detecting a weld joint sealing condition of a cover plate assembly of the spent fuel storage tank (1) and a cladding damage condition of the spent fuel assembly in the spent fuel storage tank (1), respectively; a water filling and exhausting assembly (200) connected to the spent fuel storage tank (1) and used for controllably injecting water into the spent fuel assembly for cooling and exhausting gas in the spent fuel storage tank (1); and a cutting assembly (300) used for controllably cutting a cover plate assembly weld joint of the spent fuel storage tank (1) so as to recycle the spent fuel assembly in the spent fuel storage tank (1). The present invention can implement safe recycling of the spent fuel assembly in the welded sealed storage tank.

IPC Classes  ?

• G21C 19/02 - Details of handling arrangements
• G21C 19/06 - Means for supporting or storing fuel elements or control elements

Abstract

Provided is a nuclear power plant personnel gate, which is applicable to a double-layered containment of a nuclear power plant, the double-layered containment includes outer and inner containments (1, 2), inner and outer pre-buried penetrating pieces (4, 3) are arranged on the inner and outer containments (2, 1) correspondingly, the personnel gate penetrates through the inner and outer pre-buried penetrating pieces (4, 3), the inner pre-buried penetrating piece (4) is fixedly connected with the personnel gate; the outer pre-buried penetrating piece (3) is flexibly connected with the personnel gate in a sealed mode, the nuclear power plant personnel gate includes: a shell ring assembly (5) for supporting the inner pre-buried penetrating piece (4) and the outer pre-buried penetrating piece (3); an outer sealing door (6) for establishing a shielding sealing barrier on the boundary of the outer containment (1); an inner sealing door (7) for establishing a pressure-bearing sealing barrier on the boundary of the inner containment (2); the shell ring assembly (5), the outer sealing door (6), and the inner sealing door (7) are enclosed to form a sealed space (8); and an intermediate transmission cartridge (9), which is detachably disposed in the sealed space (8), can be movably connected with the inner and outer sealing doors (7, 6), and is used for controlling opening and closing of the inner and outer sealing doors (7, 6). The reliability of the personnel gate of the nuclear power plant is improved.

IPC Classes  ?

• G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details
• G21C 13/028 - Seals, e.g. for pressure vessels or containment vessels

Abstract

A thermal-hydraulic three-dimensional real-time dynamic simulation system and method for a nuclear power plant. The thermal-hydraulic three-dimensional real-time dynamic simulation system for a nuclear power plant comprises: a modeling module (100), which is used for constructing a thermal-hydraulic three-dimensional model for a nuclear power plant, wherein the thermal-hydraulic three-dimensional model for the nuclear power plant may correspondingly control different characteristics changes by means of different node parameters; a data processing module (200), which is connected to a nuclear power plant simulator (10) and which is used to collect real-time data within the nuclear power plant simulator (10) in real time, and send the real-time data to a data driving module (300); and the data driving module (300), which is connected to the modeling module (100) and the data processing module (200) and which is used for receiving the real-time data and associating the received real-time data with the thermal-hydraulic three-dimensional model for the nuclear power plant by means of the node parameters so as to dynamically simulate the thermal-hydraulic three-dimensional model for the nuclear power plant in real time. A more intuitive and vivid reproduction of the operating conditions of the system under various working conditions of the power plant is achieved, and an intuitive and vivid three-dimensional visualization of a nuclear power thermal-hydraulic model is provided for a user.

IPC Classes  ?

• G06F 17/00 - Digital computing or data processing equipment or methods, specially adapted for specific functions
• G09B 9/00 - Simulators for teaching or training purposes

Abstract

A nuclear power plant containment (1) filtering and discharging system, comprising: a discharging pipeline assembly (2), and a first-stage filtering unit (4) and a second-stage filtering unit (5) provided on the discharging pipeline assembly (2). One end of the discharging pipeline assembly (2) is connected to a containment (1) and used for controllably conveying discharged gas in the containment (1) to the first-stage filtering unit (4); the first-stage filtering unit (4) is used for carrying out water washing filtration on the discharged gas, and carrying out hydrogen removal on the discharged gas subjected to the water washing filtration; the second-stage filtering unit (5) is connected in series to the first-stage filtering unit (4) and used for carrying out methyl iodide filtration on the discharged gas subjected to the hydrogen removal, and discharging the filtered discharged gas which meets the requirements to the atmosphere. The system effectively eliminates hydrogen in discharged gas while ensures effective retention and blocking to the discharged gas before being discharged to the atmospheric environment, so that the safety of the filtration and discharge process is improved.

IPC Classes  ?

• G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details
• G21C 13/10 - Means for preventing contamination in event of leakage

Abstract

Provided are a remote shutdown station system of a nuclear power plant, and a method for putting same into use. The system comprises: a remote shutdown station (10), an operator workstation module (100) arranged in the remote shutdown station (10), and switching unit assemblies (200) arranged inside and outside the remote shutdown station (10), wherein the operator workstation module (100) is used for receiving and processing, when a main control room is not habitable, monitoring information identical to that of the main control room; the switching unit assemblies (200) are connected to a logic control cabinet and are used for sending switching instructions according to the different states of a nuclear power plant; the logic control cabinet is in communication connection with the main control room and the operator workstation module (100), and is used for receiving the switching instructions, performing logical processing on the switching instructions, and then switching to the main control room or the operator workstation module (100) to receive and process monitoring information in a DCS monitoring network of the nuclear power plant. According to the system and the method for putting same into use, the problem that some process system logic has to be forced to be automatic when switching to the remote shutdown station system due to limited monitoring means is solved, thereby enlarging the monitoring range of a monitoring function.

IPC Classes  ?

• G21D 3/06 - Safety arrangements responsive to faults within the plant

Abstract

A safety-level functional control system and method for a steam atmospheric emissions system of a nuclear power plant. The safety-level functional control system comprises: three independent control columns, wherein each MSRIV is connected to a first control sub-column in the three independent control columns, each MSRCV is connected to a second control sub-column in one independent control column, and different MSRCVs are connected to different second control sub-columns. While meeting the triple redundant configuration of the steam atmospheric emissions system, the safety-level functional control system also meets the design principles of the three independent control columns, which simplifies the number of control columns of the control system, and reduces the complexity of the structure of the control system. The control function has a more centralized design, and the failure of any independent control column will at most lead to the loss of an opening function of one emissions loop in the steam atmosphere emissions system, and an isolation function of the steam atmosphere emissions system will remain unaffected and be highly reliable.

IPC Classes  ?

• F01D 17/10 - Final actuators
• G21D 3/04 - Safety arrangements

Abstract

A nuclear power plant equipment gate connection fastening structure and fastening method therefor. The connection fastening structure comprises a sleeve flange (10) fixedly connected to a penetrating sleeve (12) pre-embedded in a containment wall body, and a shell cover flange (20) fixedly connected to a gate shell cover (22); multiple "U"-shaped holes (30) having openings facing outward are provided at the corresponding positions of the sleeve flange (10) and the shell cover flange (20) along the circumference; after the sleeve flange (10) and the shell cover flange (20) are in butt joint, the sleeve flange (10) and the shell cover flange (20) are fastened by means of an eyelet bolt component (40); the eyelet bolt component (40) comprises a detachable eyelet bolt (42), an external nut (44), and an adjustable gasket (46) sheathed onto the eyelet bolt (42). The fastening method comprises: 1) arranging orientations according to the eyelet bolt components (40) of an equipment gate, and performing bolt pre-tightening by a matched bolt tensioner, the bolt tensioner containing 2-4 combination tension heads, each combination tension head containing 2-4 tension cylinders for pre-tightening 4-16 eyelet bolt components (40) in each time; 2) during pre-tightening, symmetrically and evenly placing the combination tension heads on the eyelet bolt components (40) which are circumferentially arranged, and connecting a hydraulic pump station (83) by means of a hydraulic pipeline (82); 3) performing tension and pre-tightening operations on the eyelet bolts (42) by using hydraulic power assist after starting the hydraulic pump station (83); 4) after one set is pre-tightened in place, performing the pre-tightening operation by switching to the next set of the eyelet bolt components (40) clockwise or counterclockwise. The connection fastening structure and fastening method therefor can enable the equipment gate to be quickly opened and closed, so as to avoid some of bolts from being not mounted due to the deformation of the sleeve flange (10) and a bolt hole of the equipment from being damaged.

IPC Classes  ?

• G21C 13/028 - Seals, e.g. for pressure vessels or containment vessels

Abstract

Provided are a multiple sequence safety display and control touch screen system of a nuclear power plant and a control method. The system comprises a non-safety processing unit for sending a sequence invoking signal, multiple single sequence safety processing units for receiving an invoking signal, and a screen switching module for switching device state information and an operation interface of a single sequence safety processing unit of an invoked sequence under a first or second screen switching mode and locking device state information and operation interfaces of other sequences. The present invention overcomes the defects in the prior art of a low working efficiency, and the increase in an operation load and a safe running risk since each sequence safety processing unit needs to be correspondingly provided with a single sequence safety display and control touch screen. By adding a screen switching module, device state information and operation interfaces of multiple single sequence safety processing units are switched to and displayed on the same M-SDCTS, thereby optimizing the design of a digital human-machine interface of a control room of a nuclear power plant and improving the working efficiency and the safety of the running of the nuclear power plant.

IPC Classes  ?

• G06F 3/0488 - Interaction techniques based on graphical user interfaces [GUI] using specific features provided by the input device, e.g. functions controlled by the rotation of a mouse with dual sensing arrangements, or of the nature of the input device, e.g. tap gestures based on pressure sensed by a digitiser using a touch-screen or digitiser, e.g. input of commands through traced gestures

Abstract

A failure downgrade running method and system for a nuclear power plant control room. The method comprises: using a KIC system to construct a pre-defense line in a control room defense in depth system (S11); using an ACP and an ECP to construct a main defense line in the control room defense in depth system (S12); using a DHP console to construct a diversity defense line in the control room defense in depth system (S13); and using an SAP console to construct a severe accident mitigation line in the control room defense in depth system (S14). Positioning of a device belonging to a severe accident defense line in a failure downgrade running policy is considered, the problem of the mixed arrangement of devices at different defense in depth levels and in different failure running modes is solved, the failure downgrade running policy of the control room is clear, and the related dedicated devices are configured so as to significantly reduce the difficulty that an operator handles failures of HSIs, reduce the human error probability, and help improving the security of a power plant.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G21D 3/06 - Safety arrangements responsive to faults within the plant
• G21D 3/04 - Safety arrangements
• G21D 3/00 - Control of nuclear power plant

Abstract

A nuclear power plant personnel airlock buffering device, comprising: a chuck structural member (12), fixed on a door frame (20) and comprising an electromagnetic chuck (120) and a chuck base plate (122) which are fixedly connected, sliding plates (124) being provided on the two sides of the chuck base plate (122), the electromagnetic chuck (120) and the chuck base plate (122) connected together being capable of moving up and down along the sliding plates (124), and the electromagnetic chuck (120) being connected to a power supply for controlling the power supply time of the electromagnetic chuck (120); and a suction plate structural member (14), fixed on a door plate (30) and comprising a suction plate (140), a suction plate connecting plate (142), and a door plate connecting plate (144), the suction plate (140) being fixed on the suction plate connecting plate (142), and a first spring (146) being provided between the suction plate connecting plate (142) and the door plate connecting plate (144); a first mandrel (148) penetrates through the interior of the first spring (146); one end of the first mandrel (148) is fixed to the suction plate connecting plate (142), and the other end penetrates through the door plate connecting plate (144), and then the first spring (146) is provided between the suction plate connecting plate (142) and the door plate connecting plate (144) by means of an auxiliary member; the chuck structural member (12) and the suction plate structural member (14) can be attracted to be locked or opened.

IPC Classes  ?

• E05F 5/02 - Braking devices, e.g. checksStopsBuffers specially for preventing the slamming of wings
• E05C 19/16 - Devices holding the wing by magnetic or electromagnetic attraction

Abstract

A support plate for a U-shaped heat transfer tube of a steam generator of a nuclear power plant, comprising a plate body (10). The plate body (10) is opened thereon with a plurality of three-leaf plum blossom-shaped holes (12) that penetrate the plate body (10). A hole bridge (14) is formed between two adjacent three-leaf plum blossom-shaped holes (12). The three-leaf plum blossom-shaped holes (12) comprise boss edges (120), side edges (122) and top edges (124). After U-shaped heat transfer tubes (16) are inserted into the three-leaf plum blossom-shaped holes (12), the boss edges (120) support tube walls of the U-shaped heat transfer tubes (16). The side edges (122) and the top edges (124) of the three-leaf plum blossom-shaped holes (12) and the tube walls of the U-shaped heat transfer tubes (16) enclose heat exchange fluid passages (18) at outer sides of the U-shaped heat transfer tubes (16). A tube gallery area (19) is formed at a middle portion of the plate body (10), the tube gallery area (19) is opened with a through hole (190), and the three-leaf plum blossom-shaped holes (12) are disposed at two sides of the tube gallery area (19), wherein the included angle of connection lines of centers of the three-leaf plum blossom-shaped holes (12) and side edges (122) corresponding thereto is α, and 0 <α≤ 13°. The support plate makes the U-shaped heat transfer tubes (16) more compactly arranged, and more U-shaped heat transfer tubes (16) may be arranged on the same area, such that the overall size and cost of the device may be reduced. The three-leaf plum blossom-shaped holes (12) have large heat exchange fluid flow areas, ensuring the structural continuity and strength at the weakest part of the hole bridges (14) of the support plate.

IPC Classes  ?

• F28F 9/013 - Auxiliary supports for elements for tubes or tube-assemblies

Abstract

A mobile emergency cooling apparatus (1) for a spent fuel pool of a nuclear power plant, comprising a power device (10), a circulating cooling pump (12), and a cooling device (14). The power device (10), the circulating cooling pump (12), and the cooling device (14) are connected in sequence, and are provided on a same mobile carrying platform (15). The inlet of the circulating cooling pump (12) is connected, by means of an inlet pipeline (16), to a water guide pipeline (18) connected to a water outlet of a primary cooling system (3) of a spent fuel pool (2). The outlet of the circulating cooling pump (12) is connected to the inlet of the cooling device (14). The outlet of the cooling device (14) is connected to a water return pipeline (22) by means of an outlet pipeline (20) to return cooled pool water to the spent fuel pool (2). A backflow pipeline (24) is connected between the outlet pipeline (20) and the inlet pipeline (16). Compared with the prior art, the mobile emergency cooling apparatus (1) for a spent fuel pool of a nuclear power plant integrates the power device (10), the circulating cooling pump (12), and the cooling device (14) onto the mobile carrying platform (15), the device integrity and usability of the mobile emergency cooling apparatus (1) can be improved. By connecting the backflow pipeline (24) between the outlet pipeline (20) and the inlet pipeline (16), the cooling device (14) is not easy to foul and implements a good cooling effect.

IPC Classes  ?

• G21D 1/02 - Arrangements of auxiliary equipment
• G21F 9/34 - Disposal of solid waste

Abstract

A non-destructive assessment method for the radiation damage of a reactor pressure vessel in a nuclear power plant, comprising the following steps: S01, monitoring in real time magnetic performance parameters of a certain monitored part of the steel of a reactor pressure vessel during a normal operation period of a nuclear power plant, the magnetic performance parameters being any one from among magnetic susceptibility χ, residual magnetization MR and coercivity HC; S02, on the basis of the measured magnetic performance parameters, calculating neutron radiation damage fluence Φ or mechanical properties of the reactor pressure vessel; S03, using the neutron radiation damage fluence Φ or the mechanical properties as analysis input parameters, performing safety assessment or service life prediction for the structural integrity of the reactor pressure vessel during a radiation damage process. The method may achieve real-time repeated non-destructive measurement, while data is accurate and test operations are safe; the present invention may simultaneously monitor the degree of radiation damage at multiple positions of the reactor pressure vessel.

IPC Classes  ?

• G21C 17/003 - Remote inspection of vessels, e.g. pressure vessels

Abstract

A nuclear power plant spent fuel storage and transportation metal tank comprises a metal tank (1) and a fuel basket (2) detachably accommodated in the metal tank. The metal tank (1) comprises: a housing (10); a base (12) hermetically sealed and fixed at a bottom portion of the housing (10); a top cover (14) hermetically sealed and fixed at a top portion of the housing (10); and a redundant top cover (16) hermetically sealed and fixed on the housing (10). In comparison to the prior art, the nuclear power plant spent fuel storage and transportation metal tank resolves issues of safe storage of spent fuel and post-transportation of the spent fuel, realizing long-term safe storage and transportation of the spent fuel. In addition, since that the fuel basket (2) is detachably accommodated in the metal tank, the fuel basket (2) can be used for loading fuel frames having different structures, providing universality.

IPC Classes  ?

• G21F 5/008 - Containers for fuel elements
• G21F 5/06 - Details of, or accessories to, the containers
• G21C 19/06 - Means for supporting or storing fuel elements or control elements

Abstract

Disclosed is a degassing device for gaseous impurities in radioactive waste liquid of a reactor of a nuclear power plant, comprising: a degassing tower (10) used for receiving the radioactive waste liquid of the reactor of the nuclear power plant and removing the gas dissolved in the waste liquid, and the degassing tower (10) being equipped with a tower top gas outlet (102) and a tower kettle liquid outlet (104); a condenser (20) communicating with the degassing tower (10) and located at the downstream side of the degassing tower, and the condenser being used for condensing the gas extracted from the top of the degassing tower (10); a vacuum pump (30) communicating with the condenser (20) and used for pumping and compressing the gas condensed by the condenser (20) and the uncondensed water vapor and maintaining the running pressure required by the interior of the tower; and a gas-water separator (30) communicating with the vacuum pump (30) and used for separating gas from water and removing the gas. Moreover, further disclosed is a degassing method for gaseous impurities in radioactive waste liquid of a reactor of a nuclear power plant. According to the degassing device and method for gaseous impurities in radioactive waste liquid of a reactor of a nuclear power plant, negative pressure degassing is adopted, the operation is convenient, safe and reliable, and system safety can be guaranteed even if system leakage occurs.

IPC Classes  ?

• G21F 9/02 - Treating gases
• G21F 9/06 - Processing

Abstract

A method and system for evaluating a safety state of a nuclear power plant, and a nuclear power plant device. The method for evaluating a safety state of a nuclear power plant comprises: acquiring operating states of all nuclear power plant safety parameters in a pre-established nuclear power plant safety parameter logical relationship model (S101); according to a pre-set safety evaluation rule and safety levels for the nuclear power plant safety parameters, evaluating the safety levels of the operating states of all the acquired nuclear power plant safety parameters so as to acquire the safety levels of all the nuclear power plant safety parameters (S102); and carrying out logical operation processing on the acquired safety levels of all the nuclear power plant safety parameters according to a safety parameter logical relationship in the established nuclear power plant safety parameter logical relationship model so as to acquire a safety level of the overall safety state of a nuclear power plant (S103). The method improves the convenience, accuracy and reliability of nuclear power plant safety state evaluation, and improves the safety and reliability of a nuclear power plant.

IPC Classes  ?

• G06Q 10/06 - Resources, workflows, human or project managementEnterprise or organisation planningEnterprise or organisation modelling

Abstract

Disclosed is an internally installed refuelling water storage tank for a nuclear-power-plant containment. The tank comprises an inner-circulation water tank (10) located between a nuclear-power-plant reactor pit (41) and a secondary shielding wall (42), and a load-bearing floorslab (20) located above the inner-circulation water tank (10). The load-bearing floorslab (20) is provided with an inner-circulation water return hole (200). An inner-circulation retention basket (30) corresponding to the inner-circulation water return hole (200) is arranged inside the inner-circulation water tank (10). An inner-circulation backflow buffering pool (32) is provided between the inner-circulation water return hole (200) and the inner-circulation retention basket (30). An annular space (40) is provided between the secondary shielding wall (42) and a nuclear-power-plant containment (44). An outer-circulation water tank (60) is arranged below a floorslab of the annular space (40). The outer-circulation water tank (60) is in communication with the annular space (40) via an outer-circulation water return hole (90). An outer-circulation retention basket (70) corresponding to the outer-circulation water return hole (90) is arranged inside the outer-circulation water tank (60). The outer-circulation water tank (60) is in communication with the inner-circulation water tank (10) via a channel (80). By means of arranging the outer-circulation water tank (60) of the refuelling water storage tank inside the containment (44), the volume of the internal refuelling water storage tank for a containment increases, and the depth of water decreases, thereby improving the earthquake resistance capacity of a reactor pit plant. The inner-circulation backflow buffering pool (32) can significantly reduce the impact force to which the inner-circulation retention basket (30) is subjected, thereby reducing the device design difficulty and costs.

IPC Classes  ?

• G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details
• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A combined nuclear fuel cycled using method for multi-reactors relates to the field of initial core fuel management of a nuclear power plant. The method includes the following steps that combined cycled using the nuclear fuels for the multi-reactors whose fuel assemblies can be interchanged, and a part of fuel assemblies can be used in the multi-reactors in turn. The method is adapted to the reactor-types of shutdown refueling or on-power refueling such as PWR, BWR, HWR, HTGR.

IPC Classes  ?

• G21C 5/00 - Moderator or core structureSelection of materials for use as moderator
• G21C 3/00 - Reactor fuel elements or their assembliesSelection of substances for use as reactor fuel elements