Abrégé

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.

Classes IPC  ?

• G21D 3/08 - Régulation de différents paramètres dans l'installation
• G21F 9/00 - Traitement des matériaux contaminés par la radioactivitéDispositions à cet effet pour la décontamination
• G21C 17/022 - Dispositifs ou dispositions pour la surveillance du réfrigérant ou du modérateur pour la surveillance de réfrigérants ou de modérateurs liquides

Abrégé

Disclosed in the present invention is a tritium production module system in a reactor. The tritium production module system has a multi-layer structure distributed in the radial direction of the reactor, and the tritium production module system comprises a tritium production area and a circulation area, wherein the tritium production area comprises a tritium breeding material, the tritium breeding material interacting with neutrons to produce tritium elements; the tritium production module system is in fluid communication with the outside of the tritium production module system via the circulation area; the circulation area comprises a tritium pipeline, the tritium pipeline being used for circulation of a reactor coolant and a tritium-containing medium, and the tritium pipeline extending in the axial direction of the reactor; a plurality of holes are provided in a pipe wall of the tritium pipeline, and the tritium pipeline is in fluid communication with the tritium production area via the holes, thereby improving the ability and efficiency of tritium transfer in the tritium production module system.

Classes IPC  ?

• G21B 1/01 - Réacteurs nucléaires hybrides fission-fusion
• G21B 1/11 - Réacteurs de fusion thermonucléaire Détails
• G21C 19/18 - Appareils pour porter les éléments combustibles à l'aire de charge du réacteur, p. ex. depuis un emplacement de stockage

Abrégé

Disclosed are a cement cured body for radioactive waste treatment in a nuclear power plant, a curing method and a curing system. The cement cured body is prepared from the following raw materials in parts by weight: 10-20 parts of radioactive waste, 59-73 parts of dry materials, 1-3 parts of a water-reducing agent and 15-25 parts of desalted water, wherein the dry materials comprise 3-5 parts of bentonite, 55-65 parts of cement and 1-3 parts of lime. The curing method comprises: weighing and mixing the raw materials to obtain a cement slurry, and curing same to obtain the cement cured body. The cement curing system comprises a curing container, a first filling part, a second filling part, a stirring device and a conveying device for transporting the curing container, wherein the first filling part and the second filling part are both provided above the curing container and are sequentially arranged in the moving direction of the curing container, and the stirring device is provided above the second filling part. The present invention enables a reduction in the non-uniform distribution of radioactive waste in the cement cured body, and enhances the performance of the cement cured body in wrapping radionuclides.

Classes IPC  ?

• G21F 9/16 - Traitements par fixation dans un milieu solide stable
• C04B 28/02 - Compositions pour mortiers, béton ou pierre artificielle, contenant des liants inorganiques ou contenant le produit de réaction d'un liant inorganique et d'un liant organique, p. ex. contenant des ciments de polycarboxylates contenant des ciments hydrauliques autres que ceux de sulfate de calcium
• G21F 9/36 - Moyens de se débarrasser des résidus solides par empaquetageMoyens de se débarrasser des résidus solides par mise en balles

Abrégé

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.

Classes IPC  ?

• G21C 7/12 - Moyens pour amener les éléments de commande dans la position désirée
• G21C 17/104 - Mesure de la réactivité
• G21C 17/112 - Mesure de la température

Abrégé

The present application relates to a steam turbine steam-seal system and a control method therefor. The steam turbine steam-seal system comprises a main steam-seal steam supply pipe (1), an auxiliary steam intake pipe (2), a steam supply regulating valve (3), a heating device (4) and a steam supply parameter measuring device (5), wherein the auxiliary steam intake pipe (2) is connected to the main steam seal steam-supply pipe (1), the steam supply regulating valve (3) is connected between the main steam seal steam-supply pipe (1) and the auxiliary steam intake pipe (2), the heating device (4) is connected between the auxiliary steam intake pipe (2) and the steam supply regulating valve (3), and the steam supply parameter measuring device (5) is arranged between the heating device (4) and the steam supply regulating valve (3).

Classes IPC  ?

• F01D 11/04 - Prévention ou réduction des pertes internes du fluide énergétique, p. ex. entre étages par obturation non contact, p. ex. du type labyrinthe utilisant un fluide d'obturation, p. ex. de la vapeur
• F01D 11/06 - Leur commande
• F01D 25/10 - Chauffage, p. ex. réchauffage avant démarrage

Abrégé

A pressure stabilizer spray valve control apparatus and method. The control apparatus comprises: a valve adjustment control loop (10) and a valve switch control loop (20), wherein the valve adjustment control loop (10) comprises a main control module (12) and a signal conditioning module (13); and the valve switch control loop (20) comprises a priority control module (24) and a switch control module (21). The switch control module (21) and the priority control module (24) are respectively used for executing switch control and priority control; the signal conditioning module (13) is used for converting a position detection signal of a valve and then transmitting same to the main control module (12); and the main control module (12) is used for performing closed-loop adjustment control according to the position detection signal and a closed-loop adjustment instruction signal. By means of providing the valve switch control loop (20) and the valve adjustment control loop (10), full-opening control, full-closing control and closed-loop adjustment control can be performed on an electromagnetic drive valve, and stable control over the valve under various working conditions can be ensured, thereby ensuring the safe and stable operation of a nuclear power station.

Classes IPC  ?

• G05B 19/042 - Commande à programme autre que la commande numérique, c.-à-d. dans des automatismes à séquence ou dans des automates à logique utilisant des processeurs numériques

Abrégé

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.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

Abrégé

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.

Classes IPC  ?

• G06F 8/41 - Compilation

Abrégé

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.

Classes IPC  ?

• G21D 1/00 - Détails des installations à énergie nucléaire
• F17D 1/04 - Systèmes de canalisation pour gaz ou vapeurs pour la distribution du gaz
• F17D 3/01 - Dispositions pour la surveillance ou la commande des opérations de fonctionnement pour commander, signaler ou surveiller le transfert d'un produit

Abrégé

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.

Classes IPC  ?

• G06V 30/422 - Dessins techniquesCartes géographiques
• G06V 30/413 - Classification de contenu, p. ex. de textes, de photographies ou de tableaux
• G06V 30/414 - Extraction de la structure géométrique, p. ex. arborescenceDécoupage en blocs, p. ex. boîtes englobantes pour les éléments graphiques ou textuels

Abrégé

A valve set for the safety of a nuclear power plant, comprising a first valve part (1-2, 2-2, 3-2) and a second valve part (1-1, 2-1, 2-3, 2-4, 3-1). The first valve part (1-2, 2-2, 3-2) comprises a valve body portion and a valve core portion; the valve body portion is provided with a sealed cavity (1-215, 2-215, 3-215) and a valve cavity (1-24, 2-24, 3-24); the valve cavity (1-24, 2-24, 3-24) is used for communicating a primary loop coolant with a medium space; one end of the valve core portion is located in the sealed cavity (1-215, 2-215, 3-215), and the other end is located in the valve cavity (1-24, 2-24, 3-24); and the valve core portion can slide in the valve body portion under the action of the pressure of a medium entering the sealed cavity (1-215, 2-215, 3-215), so as to open or close the valve cavity (1-24, 2-24, 3-24). The second valve part (1-1, 2-1, 2-3, 2-4, 3-1) is provided with an output end and input ends; the output end is communicated with the sealed cavity (1-215, 2-215, 3-215); one of the input ends is communicated with the primary loop coolant, and the other one is communicated with a pressure source; and the second valve part can control the state of communication between the output end and the input ends, so that the output end is communicated with one of the input ends. Also provided are a primary-loop hydraulic pressure relief valve set for a nuclear power plant, a valve set for primary-loop pressure relief and discharge, and an isolation valve for a passive water injection pipeline of a nuclear power plant. According to the safety valve set, the first valve part can be automatically opened when a primary loop meets requirements and is always kept in an opened state, so that the primary loop is communicated with the medium space, and the effective isolation of the first valve part can be realized when a reactor is shut down for refueling.

Classes IPC  ?

• F16K 17/02 - Soupapes ou clapets de sûretéSoupapes ou clapets d'équilibrage ouvrant sur excès de pression d'un côtéSoupapes ou clapets de sûretéSoupapes ou clapets d'équilibrage fermant sur insuffisance de pression d'un côté

Abrégé

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.

Classes IPC  ?

• G01L 13/00 - Dispositifs ou appareils pour la mesure des différences entre plusieurs valeurs de la pression des fluides

Abrégé

Disclosed in the present invention is a heat removal system for a containment vessel, comprising a heat exchanger provided in the containment vessel, a cooling liquid tank, and a steam collecting device provided in the cooling liquid tank. The heat exchanger is communicated with a medium in the cooling liquid tank; a medium in the heat exchanger absorbs heat of the containment vessel, and is discharged into the cooling liquid tank by means of the steam collecting device; and the medium in the cooling liquid tank returns to the heat exchanger. The steam collecting device comprises steam exhaust valve assemblies corresponding to different liquid level heights of the cooling liquid tank; each steam exhaust valve assembly is configured to be automatically opened under the action of its own force when the liquid level height is lowered, and at least part of the medium entering the steam collecting device enters the cooling liquid tank by means of the steam exhaust valve assemblies. The heat removal system for the containment vessel provided by the present application avoids the problem of reduction in heat removal capacity due to lowering of a liquid level, improves the driving force and the heat exchange capacity of the system, and can eliminate the risks such as steam hammer vibration, and two-phase flow-induced vibration and flow stagnation.

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur
• G21C 15/02 - Aménagement ou disposition de passages dans lesquels la chaleur est transférée au réfrigérant, p. ex. pour la circulation du réfrigérant à travers les supports des éléments combustibles

Abrégé

Disclosed in the present invention is a pressure control valve capable of replacement, the pressure control valve comprising a shell, and a throttle valve and a check valve which are mounted in the shell, wherein the shell comprises a first shell portion and a second shell portion detachably connected to a lower end of the first shell portion; the first shell portion is internally provided with a first accommodating cavity penetrating two opposite end surfaces of the first shell portion, the second shell portion is internally provided with a second accommodating cavity penetrating two opposite end surfaces of the second shell portion, and the first accommodating cavity and the second accommodating cavity are connected and in communication with each other; and the throttle valve is limited in the first accommodating cavity, and the check valve is limited in the second accommodating cavity. The pressure control valve of the present invention is used in a nuclear main pump; the separated arrangement of the shell realizes assembly and disassembly and facilitates the replacement of the throttle valve and the check valve inside; and the throttle valve and the check valve can be miniaturized and thus require a small mounting space.

Classes IPC  ?

• F16K 15/04 - Soupapes, clapets ou valves de retenue à corps de soupapes rigides guidés en forme de sphère
• F16K 47/08 - Moyens incorporés aux soupapes pour absorber l'énergie d'un fluide pour diminuer la pression, l'organe régulateur étant distinct de l'élément de fermeture

Abrégé

Disclosed is a clog-fighting leg jetting system for an offshore self-elevating platform, which comprises spud cans; spud can upper surface systems and spud can lower surface systems separately arranged on upper and lower sides of the spud cans; a supply system; and spud can main pipes connected to the supply system; each spud can upper surface system comprises first connection pipelines, a spud can main pipe splits into a plurality of first connection pipelines at the upper surface of the spud can; each first connection pipeline is connected to a plurality of first leg jetting branch pipes; an outlet end of each first leg jetting branch pipe is connected to a first leg jetting valve; each spud can lower surface system comprises second connection pipelines, a spud can main pipe splits into a plurality of second connection pipelines at the lower surface of the spud can; each second connection pipeline is connected to a plurality of second leg jetting branch pipes; and an outlet end of each second leg jetting branch pipe is connected to a second leg jetting valve. The present clog-fighting leg jetting system for an offshore self-elevating platform can prevent serious clogging of the leg jetting system, improve work efficiency, reduce the risk of strain on the legs during pull-out, and ensure the safety of the rig.

Classes IPC  ?

• E02D 13/08 - Enlèvement des obstacles
• E02D 9/02 - Enlèvement de rideaux de palplanches, de pieux, de tubages ou d'autres coffrages par arrachage
• E02B 17/00 - Iles artificielles montées sur pilotis ou supports similaires, p. ex. plates-formes sur pieds extensiblesProcédés de construction de celles-ci

Abrégé

A mechanical calculation method and apparatus for a pipeline. The method comprises: generating, according to parameter information of each component in a pipeline to be subjected to calculation, a pipeline beam model formed by connecting a plurality of pipeline beam units; acquiring a start node and an end node and meshing information of a target housing component that requires a detailed analysis; generating a target housing component mesh model formed by joining a plurality of housing units; replacing the pipeline beam unit at the corresponding position in the pipeline beam model with the target housing component mesh model to form a mixed model; performing finite element calculation with respect to the mixed model to obtain displacement of each node in the mixed model; calculating a node internal force of the pipeline beam unit according to node displacement of the pipeline beam unit; calculating stress of each housing unit according to mesh node displacement of the target housing component mesh model; and calculating stress of the pipeline beam unit according to the node internal force of the pipeline beam unit. The method enables calculation and evaluation of a pipeline and housing components to be accomplished in one pass, thereby greatly improving the working efficiency and accuracy of pipeline design.

Classes IPC  ?

• G06F 30/23 - Optimisation, vérification ou simulation de l’objet conçu utilisant les méthodes des éléments finis [MEF] ou les méthodes à différences finies [MDF]
• G06F 30/18 - Conception de réseaux, p. ex. conception basée sur les aspects topologiques ou d’interconnexion des systèmes d’approvisionnement en eau, électricité ou gaz, de tuyauterie, de chauffage, ventilation et climatisation [CVC], ou de systèmes de câblage
• G06F 113/14 - Tuyaux
• G06F 119/14 - Analyse des forces ou optimisation des forces, p. ex. forces statiques ou dynamiques

Abrégé

A mechanical calculation method and apparatus for a pipeline having a special-shaped pipe section. The method comprises: according to parameter information of each fitting in a pipeline, generating a pipeline beam model formed by connecting a plurality of pipeline beam elements; acquiring grid division information of a special-shaped pipe section requiring detailed analysis and node information of said special-shaped pipe section in the pipeline beam model; generating a grid model of a special-shaped pipe section formed by assembling a plurality of shell elements; replacing the pipeline beam elements at corresponding positions in the pipeline beam model with the grid model of the special-shaped pipe section to form a mixture model; calculating the material density of the special-shaped pipe section, and replacing the linear density in parameter information of the special-shaped pipe section to form updated parameter information of the special-shaped pipe section; performing finite element calculation on the mixture model to obtain a displacement of each node in the mixture model; and then respectively calculating stresses of the pipeline beam elements and a target shell part. According to the method, calculation and evaluation of the pipeline and special-shaped pipe shell parts can be completed at one time, thereby greatly improving the efficiency of pipeline design and the accuracy.

Classes IPC  ?

• G06F 30/23 - Optimisation, vérification ou simulation de l’objet conçu utilisant les méthodes des éléments finis [MEF] ou les méthodes à différences finies [MDF]
• G06F 30/18 - Conception de réseaux, p. ex. conception basée sur les aspects topologiques ou d’interconnexion des systèmes d’approvisionnement en eau, électricité ou gaz, de tuyauterie, de chauffage, ventilation et climatisation [CVC], ou de systèmes de câblage
• G06F 113/14 - Tuyaux
• G06F 119/14 - Analyse des forces ou optimisation des forces, p. ex. forces statiques ou dynamiques

Abrégé

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.

Classes IPC  ?

• G06F 30/13 - Conception architecturale, p. ex. conception architecturale assistée par ordinateur [CAAO] relative à la conception de bâtiments, de ponts, de paysages, d’usines ou de routes
• G06T 17/00 - Modélisation tridimensionnelle [3D] pour infographie

Abrégé

A passive spraying system for aerosol removal in a nuclear power plant, and a containment system. The passive spraying system comprises: a condensate collector (3), a flow guide pipe (4), a spray water tank (5), a spray pipe (19), a burst valve (6) and spray nozzles (7). The condensate collector (3) is connected to the spray water tank (5) by means of the flow guide pipe (4), and the condensate collector (3) is used for collecting condensate inside a containment (1). The spray water tank (5) is connected to the spray nozzles (7) by means of the spray pipe (19). The burst valve (6) is provided on the spray pipe (19). When the pressure inside the containment (1) reaches the start-up triggering pressure of the burst valve (6), the burst valve (6) opens, such that water inside the spray water tank (5) flows into the spray nozzles (7) by means of the spray pipe (19) and the spray nozzles (7) spray water to remove aerosols inside the containment (1).

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur

Abrégé

Disclosed in the present invention are a nuclear island reactor plant construction equipment integration platform and a climbing method therefor. The nuclear island reactor plant construction equipment integration platform comprises a steel platform system, a supporting and jacking system, a hanger and formwork system, and construction integration equipment. The supporting and jacking system is mounted on the side surface of a nuclear island reactor plant. The supporting and jacking system is located below the steel platform system, and the hanger and formwork system is installed below the steel platform system. The steel platform system comprises an annular steel truss, a primary radial steel truss and a secondary radial steel truss. The construction integration equipment comprises an annular rail installed at the bottom of the hanger and formwork system and ring cranes slidably connected to the annular rail. According to the nuclear island reactor plant construction equipment integration platform, the problem that the total time of construction is restricted by the "stepped" process of an inner shell, an outer shell and a surrounding plant in the existing construction of the nuclear island reactor plant is solved.

Classes IPC  ?

• E04G 21/04 - Dispositifs pour transporter et distribuer en même temps
• E04G 11/28 - Coffrages montants, c.-à-d. coffrages qui ne sont pas en contact, pendant qu'on les élève de couche en couche, avec le béton versé
• E04G 3/18 - Échafaudages essentiellement supportés par le bâtiment, p. ex. réglables en hauteur supportés par des consoles ou d'autres dispositions montées dans des ouvertures des bâtiments, p. ex. des ouvertures de fenêtres
• E04G 3/28 - Échafaudages mobilesÉchafaudages avec plates-formes mobiles

Abrégé

A dual-energy X-ray detector structure and a dual-energy X-ray detection method. The dual-energy X-ray detector structure comprises a power source assembly and a detection assembly, wherein the detection assembly comprises a photon absorber (32), a first electrode assembly and a second electrode assembly; the photon absorber (32) comprises a first side, and a second side arranged opposite the first side; the first electrode assembly is attached to the first side, and the second electrode assembly is attached to the second side; the first electrode assembly comprises a first electrode (34) and a second electrode (33), the first electrode (34) is connected to a positive electrode output end, and the second electrode (33) is connected to a grounding output end; and the second electrode assembly comprises a third electrode (36) and a fourth electrode (35), the third electrode (36) is connected to the positive electrode output end, and the fourth electrode (35) is connected to the grounding output end. The dual-energy X-ray detector does not require calibration of time and space; the detection speed is fast, and the cost of the detector is low; and signals of a low-energy X-ray (30) and a high-energy X-ray (31) can be respectively acquired, such that the density of an object to be detected can be better distinguished.

Classes IPC  ?

• G01T 1/16 - Mesure de l'intensité de radiation

Abrégé

Disclosed are an unblocking apparatus for a furnace discharging pipe and a use method. The unblocking apparatus includes a rail, a rail car that may move along the rail, an unblocking drive mechanism arranged on the rail car, a heat-unblocking component, a cold-unblocking component, and a material receiving component that is used to receive a blocking material in the discharging pipe, and a drive end of the unblocking drive mechanism is detachably connected with one end of the heat-unblocking component and the cold-unblocking component respectively. The present application effectively handles different blockage situations of the furnace discharging pipe by connecting the unblocking drive mechanism with an unblocking rod capable of heat-unblocking and a drilling rod capable of cold-unblocking, thereby two modes of heat-unblocking and cold-unblocking are performed on the furnace discharging pipe; and the discharging pipe may be unblocked by a remote operation.

Classes IPC  ?

• F27D 25/00 - Dispositifs pour enlever les incrustations
• F27D 3/15 - Équipement de piquéeÉquipement pour l'élimination du laitier

Abrégé

The present invention relates to a method and system for analyzing and optimizing a key path in a debugging stage of a nuclear power unit. The method comprises the following steps: determining key nodes of the key path in the debugging stage of the nuclear power unit; performing key path analysis according to the key nodes to obtain a key path analysis result; and performing key path optimization according to the key path analysis result to obtain an optimized key path. According to the present invention, the key nodes of the key path in the debugging stage of the nuclear power unit are analyzed and determined, and the key path is accurately recognized, optimized and adjusted, such that an optimal key path can be obtained without using an early-stage unit as a reference, thereby ensuring ordered operation of a project.

Classes IPC  ?

• G06Q 10/04 - Prévision ou optimisation spécialement adaptées à des fins administratives ou de gestion, p. ex. programmation linéaire ou "problème d’optimisation des stocks"

Abrégé

A method for improving the sealing performance of a containment (6) of a nuclear power plant. The method comprises the following steps: 1) determining a weak point (1) where there may be a gap in a containment (6); 2) establishing an independent cover body (7) outside the weak point (1), such that the weak point (1) is in an independent space (2) of the independent cover body (7); 3) pressurizing the independent cover body (7) to a preset pressure by means of an air compressor (3); 4) introducing an aerosol source material into an aerosol generator (4), emitting aerosol particles into the independent cover body (7) by means of the aerosol generator (4), enabling the aerosol particles to permeate into the gap under the action of the preset pressure, and depositing the aerosol particles in the gap so as to seal the gap; and 5) monitoring a pressure change in the independent cover body (7) in real time, and evaluating the sealing effect on the gap, so as to complete the sealing of the gap. By using a natural deposition phenomenon of aerosol in the gap, and by means of the dynamic characteristics of the aerosol particles themselves, the aerosol particles seal the weak point (1) where there may be a gap in the containment (6), such that the leakage rate of the containment (6) is reduced.

Classes IPC  ?

• G21C 13/028 - Joints, p. ex. pour les cuves de pression ou les enceintes de confinement
• G21C 17/00 - SurveillanceTests

Abrégé

The present invention discloses a reactor accident simulation test device and method. The device comprises: a primary loop model which comprises a heating unit, at least one steam generator primary side and a pipeline which connects the heating unit and the steam generator primary side to form a primary loop circuit; a secondary loop model which comprises a steam collection device, at least one steam generator secondary side corresponding to the steam generator primary side and a pipeline which connects the steam generator secondary side and the steam collection device to form a secondary loop circuit; an accident simulation unit for simulating the occurrence of a large break loss of coolant accident and/or a main steam line break accident; an acquisition unit for acquiring target parameters; and a master control device for producing a test result on the basis of the target parameters. The test device simulates a test device of which design parameters are consistent with those of a reactor prototype by means of the primary loop model and the secondary loop model, and has the advantages of high test value and the like.

Classes IPC  ?

• G21C 17/00 - SurveillanceTests

Abrégé

A method for optimizing multiple parameters of a nuclear reactor. The method comprises: acquiring parameters to be optimized of a target optimization object in a nuclear reactor, and an initial design scheme of the nuclear reactor; generating an optimization function of the target optimization object according to the initial design scheme and said parameters; calling a parameter optimization model to solve the optimization function to obtain optimized values of said parameters, wherein the parameter optimization model is pre-constructed according to an adaptive moment estimation algorithm; determining, on the basis of the optimized values of said parameters, whether the optimization function converges, and if the optimization function converges, determining the optimized values of said parameters to be target optimized values of said parameters; and updating the initial design scheme according to the target optimized values of said parameters.

Classes IPC  ?

• G06Q 10/04 - Prévision ou optimisation spécialement adaptées à des fins administratives ou de gestion, p. ex. programmation linéaire ou "problème d’optimisation des stocks"

Abrégé

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.

Classes IPC  ?

• E02B 9/04 - Canaux à écoulement libre ou canaux d'amenéePrises d'eau

Abrégé

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).

Classes IPC  ?

• G01M 7/02 - Test de vibration
• G01M 10/00 - Tests hydrodynamiquesAménagements dans ou sur les bassins de tests des navires ou les tunnels hydrauliques

Abrégé

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.

Classes IPC  ?

• G21C 15/14 - Aménagement ou disposition de passages dans lesquels la chaleur est transférée au réfrigérant, p. ex. pour la circulation du réfrigérant à travers les supports des éléments combustibles provenant des conduits servant au passage d'un fluide chaudAménagement ou disposition de passages dans lesquels la chaleur est transférée au réfrigérant, p. ex. pour la circulation du réfrigérant à travers les supports des éléments combustibles provenant des conduits comprenant des appareils auxiliaires, p. ex. pompes, caméras
• G21C 5/10 - Moyens pour supporter la structure complète
• G21C 3/322 - Moyens pour influencer l'écoulement du réfrigérant à travers ou autour des faisceaux
• G21C 7/32 - Commande de la réaction nucléaire par variation du courant de réfrigérant à travers le cœur

Abrégé

The present application provides a pressurized water reactor control method and apparatus, a computer device and a storage medium. The method comprises: if a pressurized water reactor is eligible for a dilution operation, performing the dilution operation on the pressurized water reactor according to a first dilution flow rate; continuously monitoring the boron concentration of a primary loop coolant of the pressurized water reactor, and continuously monitoring the reciprocal of the neutron count rate of the pressurized water reactor through a source range channel of the pressurized water reactor; in the dilution operation process, adjusting the first dilution flow rate at least once according to the degree of reduction of the boron concentration of the primary loop coolant and the reciprocal of the neutron count rate; if the boron concentration of the primary loop coolant meets a first preset condition, stopping the dilution operation; and if the boron concentration of the primary loop coolant meets a second preset condition, selecting an operation mode for controlling the pressurized water reactor according to the reciprocal of the neutron count rate. The method can be used to improve the control efficiency of the pressurized water reactor and reduce the accidental critical risk.

Classes IPC  ?

• G21C 7/36 - Circuits de commande
• G21C 17/022 - Dispositifs ou dispositions pour la surveillance du réfrigérant ou du modérateur pour la surveillance de réfrigérants ou de modérateurs liquides

Abrégé

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).

Classes IPC  ?

• G21C 17/04 - Détection de rupture de gaine
• G21C 17/10 - Combinaison structurelle de l'élément combustible, de la barre de commande, du cœur du réacteur, ou de la structure du modérateur avec des instruments sensibles, p. ex. pour la mesure de la radioactivité, des contraintes

Abrégé

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).

Classes IPC  ?

• G05B 19/418 - Commande totale d'usine, c.-à-d. commande centralisée de plusieurs machines, p. ex. commande numérique directe ou distribuée [DNC], systèmes d'ateliers flexibles [FMS], systèmes de fabrication intégrés [IMS], productique [CIM]

Abrégé

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.

Classes IPC  ?

• F01D 21/12 - Arrêt des "machines" ou machines motrices, p. ex. dispositifs d'urgenceDispositifs de régulation, de commande ou de sécurité non prévus ailleurs sensibles à la température
• G01D 3/00 - Dispositions pour la mesure prévues pour les objets particuliers indiqués dans les sous-groupes du présent groupe
• G21C 17/00 - SurveillanceTests

Abrégé

Provided are a steel lining module construction method, a hoisting tool (100) and a limiting tool (130). The steel lining module construction method comprises: assembling a bottom plate (10), a first truncated cone (30) and a second truncated cone (40) together to form a steel lining module (1); arranging first hoisting points (103) on the steel lining module (1); connecting the first hoisting points (103) on the steel lining module (1) to the hoisting tool (100), and connecting the hoisting tool (100) to a hoisting apparatus (120); and hoisting the steel lining module (1) into a containment in a nuclear island reactor building. By means of the construction method, the overall assembly of the module and the civil engineering of the containment and the construction of an internal structure are independent from each other, so that cross-operation risk is reduced, thereby shortening the construction progress, and improving the construction management safety quality.

Classes IPC  ?

• B66C 1/12 - Élingues comportant chaînes, fils métalliques, cordes ou bandesFilets
• B66C 1/14 - Élingues à crocs
• B66C 13/08 - Dispositifs auxiliaires pour commander les mouvements des charges suspendues ou pour empêcher le câble de prendre du mou pour déposer les charges selon un orientement ou dans une position donnés
• B66C 13/16 - Utilisation de dispositifs indicateurs de positionnement ou de pesée
• B66C 13/18 - Systèmes ou dispositifs de commande

Abrégé

A scale determination method and apparatus for a reactor test model, a computer device, a storage medium, and a computer program product. The method comprises: determining a target reactor, and acquiring a test demand analysis result corresponding to the target reactor under a test working condition (S102); acquiring a test model to be constructed corresponding to the target reactor, and a power ratio and a volume ratio between said test model and the target reactor (S104); on the basis of the power ratio, the volume ratio, and a modular analysis mode corresponding to the test demand analysis result, determining multiple groups of modular proportion combinations corresponding to said test model (S106); performing feasibility analysis processing on each group of modular proportion combinations, and selecting a target modular proportion combination on the basis of processing results (S108); and according to the power ratio, the volume ratio and the target modular proportion combination, determining an overall scale parameter of said test model (S110).

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

Abrégé

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.

Classes IPC  ?

• G21C 19/02 - Détails des dispositions pour la manipulation
• G21C 19/19 - Parties de réacteurs spécifiquement adaptées pour faciliter la manipulation, p. ex. pour faciliter le chargement ou le déchargement des éléments combustibles
• G21C 19/20 - Dispositions pour introduire des objets à l'intérieur de l'enceinte sous pressionDispositions pour manipuler des objets à l'intérieur de l'enceinte sous pressionDispositions pour extraire des objets de l'enceinte sous pression
• G21C 17/00 - SurveillanceTests
• G21D 1/02 - Dispositions d'équipement auxiliaire

Abrégé

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.

Classes IPC  ?

• C02F 9/00 - Traitement en plusieurs étapes de l'eau, des eaux résiduaires ou des eaux d'égout

Abrégé

A stepwise pressure relief device, comprising a steam intake pipeline, a silencing module and a heat recovery module. The steam intake pipeline comprises a first pipeline and at least two discharge pipelines, wherein the first pipeline is in communication with the silencing module and the heat recovery module, and the at least two discharge pipelines are respectively in communication with both the heat recovery module and the atmosphere. The silencing module is in communication with the heat recovery module and the atmosphere, and the heat recovery module is in communication with the atmosphere and users. A steam generator secondary-loop heat sink system, which has a steam delivery and pressure relief pipeline containing a stepwise pressure relief device. By means of the stepwise pressure relief device and the steam generator secondary-loop heat sink system, the economy, and the economy and safety of a power plant can be improved.

Classes IPC  ?

• G21D 3/06 - Dispositions de sécurité réagissant à des défaillances à l'intérieur de l'installation
• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur

Abrégé

A performance experiment system for a closed Brayton cycle, comprising a main cycle system (2), a gas source system, a heat source system (1) and a power system (7). The main cycle system (2) comprises a turbine (201), a gas compressor (202) and a cooling unit. The inlet of the gas compressor (202) is connected to the gas source system so as to feed an experimental gas and compress same, and the outlet of the gas compressor (202) is connected to the heat source system (1) so as to heat the compressed experimental gas. The inlet of the turbine (201) is connected to the heat source system (1) so as to do work on the heated experimental gas. The cooling unit is separately connected to the turbine (201) and the gas compressor (202) and used for cooling exhaust gas after work and returning same to the gas compressor (202). The power system (7) comprises an integrated starter generator (203), the integrated starter generator (203) being connected to the gas compressor (202) and used for driving the gas compressor (202) to operate so as to perform compression and generating power when the output work of the turbine (201) is greater than the consumed work of the gas compressor (202). The experiment system can carry out various system-level experiments for verifying loop flow characteristics, heat transfer characteristics, mechanical characteristics, transmission characteristics, etc.

Classes IPC  ?

• G01M 15/05 - Test des moteurs à combustion interne par contrôle combiné d'au moins deux paramètres différents des moteurs
• G01M 15/02 - Détails ou accessoires pour appareils de test
• G01M 15/00 - Test des moteurs

Abrégé

A test piece and a manufacturing method therefor. The test piece comprises a test piece substrate and a defect unit. The defect unit comprises a platform (8) and an attachment (9), wherein the platform (8) is arranged in a groove of the test piece substrate, and a first artificial defect (6) having a predetermined shape and/or size is either machined in the platform (8) or not. When the first artificial defect (6) is machined in the platform (8), a second artificial defect (7) having a predetermined shape and/or size corresponding to those of the first artificial defect (6) in the platform (8) is machined in the attachment (9), which is assembled with the platform (8) in a fitted manner, and the first artificial defect (6) and the second artificial defect (7) are assembled to form a complete artificial defect. Alternatively, when the first artificial defect (6) is not machined in the platform (8), a second artificial defect (7) having a predetermined shape and/or size is machined in the attachment (9), which is assembled with the platform (8) in a fitted manner, and the second artificial defect (7) alone forms a complete artificial defect. After the groove of the test piece substrate is subjected to filling welding, a test piece containing, at a contact interface between the platform (8) and the attachment (9), a high-precision artificial defect having a predetermined shape, size and position can be manufactured.

Classes IPC  ?

• G01N 1/28 - Préparation d'échantillons pour l'analyse
• G01N 23/02 - Recherche ou analyse des matériaux par l'utilisation de rayonnement [ondes ou particules], p. ex. rayons X ou neutrons, non couvertes par les groupes , ou en transmettant la radiation à travers le matériau
• G01N 23/04 - Recherche ou analyse des matériaux par l'utilisation de rayonnement [ondes ou particules], p. ex. rayons X ou neutrons, non couvertes par les groupes , ou en transmettant la radiation à travers le matériau et formant des images des matériaux
• G01N 29/04 - Analyse de solides

Abrégé

Disclosed in the present invention are a risk prediction method and prediction system for a nuclear power unit, and a risk assessment system for same. The method comprises: acquiring a function of the performance state of each device/part of a nuclear power unit varying along with time; according to the function of the performance state varying along with time, updating a risk prediction parameter value of a preset living probability safety analysis (living PSA) model at a preset future moment; determining whether configurations of the nuclear power unit is changed at the preset future moment, and if so, updating the structure of the living PSA model according to the changed configurations; and applying the risk prediction parameter value to the living PSA model of the latest structure, so as to obtain a risk prediction value of the nuclear power unit at the preset future moment. The present invention can realize risk prediction for a nuclear power unit, and improve the accuracy of a risk prediction value of the nuclear power unit, such that a risk of the nuclear power unit at a future moment is foreseen.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu
• G06F 119/02 - Analyse de fiabilité ou optimisation de fiabilitéAnalyse de défaillance, p. ex. performance dans le pire scénario, analyse du mode de défaillance et de ses effets [FMEA]

Abrégé

The present invention relates to the technical field of heat exchange devices, and relates to a wall having functions of noise elimination, phase change cold storage, and heat transfer enhancement. The wall comprises a concrete layer (1), a phase change cold storage layer (2), a micro-perforated plate (5), a heat insulation layer (8), and a heat-conduction cold bridge. The wall having functions of noise elimination, phase change cold storage, and heat transfer enhancement of the present invention can implement the functions of noise elimination, phase change cold storage, and heat transfer enhancement when being applied to nuclear power station master control rooms and electrical instrument device rooms.

Classes IPC  ?

• E04B 2/00 - Murs, p. ex. cloisons, pour bâtimentsStructure des murs en ce qui concerne l'isolationAssemblages spécifiques pour les murs
• E04B 1/88 - Éléments à la fois isolants acoustiques et thermiques
• E04B 1/94 - Protection contre d'autres agents indésirables ou dangers contre le feu

Abrégé

A gas-cooled micro-reactor straight-pipe-type main pipeline, which is used for connecting a reactor system (1) and a power generation system (3). The main pipeline (2) comprises an outer pipeline (6) and an inner pipeline (8), wherein the inner pipeline (8) is fixed to the outer pipeline (6) by means of several supporting members (14), which are uniformly arranged in a circumferential direction; the inner pipeline (8) sequentially comprises a housing (20), an intermediate heat-insulating layer (21) and a lining (22), which are coaxially arranged from outside to inside; and heat displacement compensation mechanisms are respectively arranged on the housing (20) and the lining (22) of the inner pipeline (8). The straight-pipe-type main pipeline employs the design solution of an inner pipe and an outer pipe, and key technologies such as heat insulation and heat compensation, thereby saving on arrangement space and having the effect of heat preservation and heat insulation.

Classes IPC  ?

• F16L 9/19 - Tuyaux à canaux multiples ou assemblages de tuyaux
• F16L 59/14 - Dispositions pour l'isolation des tuyaux ou des tuyauteries
• F16L 51/00 - Compensation de la dilatation dans les canalisations
• F16L 51/03 - Compensation de la dilatation dans les canalisations utilisant un soufflet ou un tube dilatable en accordéon ou ondulé comprenant plusieurs soufflets
• F16L 27/12 - Raccords réglablesRaccords permettant un déplacement des parties raccordées permettant réglage ou déplacement sensiblement longitudinal
• F16L 39/00 - Raccords ou accessoires de raccordement pour tuyaux à double paroi ou à canaux multiples ou pour assemblages de tuyaux
• G21D 1/00 - Détails des installations à énergie nucléaire

Abrégé

A security defense system and method for industrial control systems, relating to the technical field of network security of industrial control systems of nuclear power plants. A first industrial control system for achieving the most important function in the industrial control systems is located in the center of the defense system and has minimum network interfaces with other industrial control systems, and data is unidirectionally transmitted outwards. A second and a third industrial control system for achieving a secondary function receive network data of the first industrial control system, and data is unidirectionally transmitted to a fourth industrial control system. The fourth industrial control system for achieving a non-important function does not have a network interface to the second industrial control system and the third industrial control system, and data is unidirectionally transmitted outwards to an external system. The defense system comprises, from outside to inside, a boundary protection layer, a network communication protection layer, and a host protection layer. By means of the defense in depth system of the overall architecture of the industrial control systems, network security protection is carried out on a hierarchical basis according to the point of focus, and the network security protection capabilities of the industrial control systems of the nuclear power plant can be improved.

Classes IPC  ?

• H04L 9/40 - Protocoles réseaux de sécurité

Abrégé

A water level control system of a pressure regulator of a pressurized water reactor comprises a heat exchanger, a one-loop system, a temperature measurement module, a water level setting value generation module, a pressure regulator, a water level measurement apparatus, a rotational speed control module, and an upper filling pump (5). The heat exchanger communicates with the one-loop system by means of an upper filling channel (1) and a lower discharge channel (2), respectively. The temperature measurement module respectively measures the temperature of a cold pipe section (3) and a hot pipe section (4) of the one-loop system and obtains the average temperature of the one-loop system. The temperature measurement module is in communication connection with the water level setting value generation module, and the water level setting value generation module, according to the average temperature, generates a water level setting value of the pressure regulator. The water level measurement apparatus measures the real-time water level value of the pressure regulator. The rotational speed control module, according to the water level setting value and the real-time water level value, generates a control command for controlling the rotational speed of the upper filling pump (5). The upper filling pump (5) is arranged in the upper filling channel, so as to supply water to the one-loop system, and the upper filling pump (5) controls the water supply speed according to the control command, such that adjustment precision can be ensured while frequent actions of an actuating mechanism are avoided.

Classes IPC  ?

• G21D 3/04 - Dispositions de sécurité

Abrégé

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.

Classes IPC  ?

• G21D 3/00 - Commande des installations à énergie nucléaire

Abrégé

The present invention provides a nuclear-grade pipeline support arrangement method and apparatus. The method comprises: acquiring a target zone where supports can be arranged on pipelines; in the target zone, selecting a position where a pipeline has a maximum displacement under seismic conditions to arrange a damper step by step, until the seismic stress of the pipelines is small enough to meet standard requirements of the seismic conditions; and replacing the damper having a minimum displacement under thermal expansion conditions among all the dampers with a rigid support step by step, until the maximum thermal expansion stress of the pipelines does not exceed standard requirements or there is no remaining damper, so as to obtain a target support arrangement scheme. According to the method, the problems in the existing nuclear-grade pipeline support arrangement methods that blindly arranging expensive dampers takes a long time and lacks systematicness, and it is difficult to achieve optimal selection of pipe support types can be solved.

Classes IPC  ?

• F16L 55/02 - Absorbeurs d'énergieAbsorbeurs de bruit
• F16L 55/035 - Absorbeurs de bruit sous forme de potences ou de supports spécialement adaptés

Abrégé

A power distribution measurement method, apparatus and system for a nuclear power plant. The measurement method comprises: acquiring a value of initial reactivity of a reactor when a control rod is adjusted from the inside of a reactor core of the reactor to the top of the reactor core (S100); acquiring a value of final reactivity of the reactor when the control rod is adjusted from the inside of the reactor core to the bottom of the reactor core (S200); and acquiring a radial power distribution of the reactor core by means of the value of the initial reactivity, the value of the final reactivity, and a theoretical power of a control rod assembly corresponding to the control rod (S300). Compared with traditional technology, the method may be used to realize a power distribution measurement of a reactor core in a zero-power physical test platform, thereby improving the accuracy of the power distribution measurement in the zero-power physical test platform.

Classes IPC  ?

• G21C 17/10 - Combinaison structurelle de l'élément combustible, de la barre de commande, du cœur du réacteur, ou de la structure du modérateur avec des instruments sensibles, p. ex. pour la mesure de la radioactivité, des contraintes
• G21C 17/104 - Mesure de la réactivité

Abrégé

The present invention provides a validity determination method for a target signal, a processing method, a device, and a decision-making system. The validity determination method for the target signal comprises: determining whether a target signal has a quality bit; if yes, further determining whether the value of the quality bit is a preset first numerical value, wherein the first numerical value indicates that the state of the quality bit is good; if the value of the quality bit is the preset first numerical value, further determining, according to a self-diagnosis function related to the target signal in a decision-making system, whether the initial validity of the target signal is valid; and if the initial validity of the target signal is invalid, determining that the final validity of the target signal is invalid. The method, the device, and the decision-making system can solve the problem in the related art that the validity determination of a signal cannot be realized because no validity determination method for the target signal of the decision-making system exists.

Classes IPC  ?

• G06Q 10/06 - Ressources, gestion de tâches, des ressources humaines ou de projetsPlanification d’entreprise ou d’organisationModélisation d’entreprise ou d’organisation

Abrégé

Provided are a pipeline support position optimization method and a device. The method comprises: obtaining all support positions in a pipeline and determining whether each support is a movable support; configuring a pipeline support position optimization target, performing coordinated adjustment on all the movable support positions to meet the pipeline support position optimization target, and obtaining optimized pipeline support positions. The method and apparatus can simultaneously adjust the positions of multiple supports, and have high optimization efficiency.

Classes IPC  ?

• G06F 30/18 - Conception de réseaux, p. ex. conception basée sur les aspects topologiques ou d’interconnexion des systèmes d’approvisionnement en eau, électricité ou gaz, de tuyauterie, de chauffage, ventilation et climatisation [CVC], ou de systèmes de câblage
• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu
• G06F 111/04 - CAO basée sur les contraintes

Abrégé

A digital twin technology-based containment twin system and a construction method therefor. A digital mapping body corresponding to a physical entity containment system is constructed in a virtual space of a computer or a server, and the digital mapping body comprises a virtual reality simulation module, a multi-physical field coupling calculation module, a durability and service life prediction calculation module, a machine learning prediction calculation module, and a feedback control and warning module. Cognition and data collection are performed by means of a sensor system and a data feedback and collection system placed in a physical entity containment so as to form a cognitive database of the physical entity containment system, and the cognitive database is used as a data input source of the digital mapping body. The digital mapping body performs simulation calculation by means of each calculation module, and transmits information to each related party, so that an information acquisition channel of each related party is unblocked and an information exchange channel between the related parties is opened, thereby better increasing the capability of responding to an accident in a nuclear power plant, and improving the safety of nuclear power.

Classes IPC  ?

• G06F 30/27 - Optimisation, vérification ou simulation de l’objet conçu utilisant l’apprentissage automatique, p. ex. l’intelligence artificielle, les réseaux neuronaux, les machines à support de vecteur [MSV] ou l’apprentissage d’un modèle
• G06F 119/04 - Analyse de vieillissement ou optimisation contre le vieillissement
• G06F 119/08 - Analyse thermique ou optimisation thermique
• G06F 119/14 - Analyse des forces ou optimisation des forces, p. ex. forces statiques ou dynamiques

Abrégé

A reactor online protection method and system, and a readable storage medium. The reactor online protection method comprises: acquiring n groups of detection data of a self-powered detector (S10); taking m groups of different detection data from among the n groups of detection data, so as to form a subset, and respectively reconstructing reactor core power distribution data of the corresponding subset according to the detection data of each subset (S20); according to the reactor core power distribution data corresponding to each subset, respectively calculating a maximum LPD value and a minimum DNBR value which correspond to each subset (S30); selecting a final maximum LPD value from among the maximum LPD values corresponding to the subsets, and selecting a final minimum DNBR value from among the minimum DNBR values corresponding to the subsets (S40); and according to the final maximum LPD value, the final minimum DNBR value, a preset LPD setting value and a preset DNBR setting value, determining whether to trigger a reactor trip signal (S50).

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

Abrégé

The present invention relates to the field of artificial intelligence, and specifically relates to a real-time target tracking method, a device, and a storage medium. The method comprises: selecting image data required for training, and enhancing a search image frame; configuring a feature extraction network and an RPN network, and performing regression on a bounding box by the RPN network; removing the RPN network, carrying out stage one training of a backbone network, adding an RPN network posterior to the backbone network, and carrying out stage two training; for an unknown video stream, marking a target to be tracked, and performing training to obtain a specialized network; taking each frame of the video stream as a search image frame, inputting the search image frames into the specialized network, performing regression to output bounding boxes, and completing target tracking. The present invention implements tracking of a target in a video on the basis of a siamese network; also, with respect to the problems of large amounts of memory being used and low inference speed that are pervasive in existing siamese networks, rapid target tracking is achieved without an excessive loss of precision by means of a network trained in the present invention.

Classes IPC  ?

• G06V 10/77 - Traitement des caractéristiques d’images ou de vidéos dans les espaces de caractéristiquesDispositions pour la reconnaissance ou la compréhension d’images ou de vidéos utilisant la reconnaissance de formes ou l’apprentissage automatique utilisant l’intégration et la réduction de données, p. ex. analyse en composantes principales [PCA] ou analyse en composantes indépendantes [ ICA] ou cartes auto-organisatrices [SOM]Séparation aveugle de source
• G06K 9/62 - Méthodes ou dispositions pour la reconnaissance utilisant des moyens électroniques

Abrégé

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).

Classes IPC  ?

• G06F 30/17 - Conception mécanique paramétrique ou variationnelle
• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu
• G06F 17/00 - Équipement ou méthodes de traitement de données ou de calcul numérique, spécialement adaptés à des fonctions spécifiques

Abrégé

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.

Classes IPC  ?

• G21C 13/032 - Raccords entre le tube et la paroi d'une enceinte, p. ex. tenant compte des contraintes thermiques
• G21C 13/06 - Bouchons scellés

Abrégé

The present application relates to a reactor core power online monitoring method and system. The reactor core power online monitoring method comprises: obtaining measurement data of each self-powered detector in each group, wherein self-powered detectors in a reactor core are divided into 2-4 groups, a plurality of self-powered detectors in each group are uniformly distributed on the reactor core in the radial direction of the reactor core and are pairwise centrosymmetric relative to a reactor core original point, and the plurality of self-powered detectors of each group permeate each other in an intersected manner; and respectively determining, according to the measurement data, whether a failure occurs in each self-powered detector in each group, and for a self-powered detector in which a failure occurs, taking measurement data of a self-powered detector which is centrosymmetric with the self-powered detector as own measurement data.

Classes IPC  ?

• G21C 17/108 - Mesure du flux

Abrégé

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.

Classes IPC  ?

• G21C 7/12 - Moyens pour amener les éléments de commande dans la position désirée
• G21C 19/10 - Dispositifs de relèvement ou d'enlèvement adaptés pour coopérer avec les éléments combustibles ou avec l'élément de commande

Abrégé

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.

Classes IPC  ?

• G06F 9/445 - Chargement ou démarrage de programme

Abrégé

The present application relates to a reactor modeling data acquisition method and apparatus, and a device, a medium and a program product. The method comprises: performing hierarchical division processing on apparatuses which are contained in each working process that is involved in a target reactor, so as to obtain a hierarchical division result corresponding to each working process; performing control volume division on apparatuses which are contained in the lowest hierarchy in the hierarchical division result, so as to obtain an initial division result, wherein the initial division result comprises a plurality of first control volumes; for each initial division result, performing correction processing on the initial division result according to a physical phenomenon identification table, so as to obtain a target division result, wherein the target division result comprises a plurality of second control volumes, and each second control volume comprises at least one apparatus having the same structure and being related to the same physical phenomenon; and generating, according to each second control volume and a physical phenomenon related to the second control volume, modeling data corresponding to the target reactor. By means of the present method, the physical phenomena of a reactor can be comprehensively identified.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

Abrégé

The present invention belongs to the technical field of welding, and particularly disclosed are welding systems and welding control methods. The welding systems and the welding methods are used for welding a first pipe fitting to a structural member, and a groove is formed in the structural member. One welding system comprises a welding robot, a measuring device and a control device. After the first pipe fitting and the structural member are pre-positioned, the measuring device is used for measuring first position information of the first pipe fitting and a first contour of the groove. The control device is configured to establish a groove model according to the first position information and the first contour, calculate to obtain actual coaxiality data of the first pipe fitting, and according to the groove model and the actual coaxiality data, analyze in combination with the target coaxiality data so as to determine a first welding process. The control device is also configured to analyze, on the basis of the groove model and the target coaxiality, so as to determine the first welding process, thus the welding robot can perform efficient automatic welding of the first pipe fitting and the structural member, the deformation amount after the pipe fitting is welded is reduced, and the welding precision is improved.

Classes IPC  ?

• B23K 9/127 - Moyens pour suivre des lignes au cours du soudage ou du découpage à l'arc

Abrégé

Disclosed in the present invention are a method and apparatus for generating an association viewing model. The method comprises: acquiring a data object; calling a preset target character extraction script to perform an extraction operation for the data object, so as to obtain a target character code; and establishing an association relationship between the target character code and the data object, and then generating an association viewing model. By establishing an association relationship between a coding object and various data objects, data query, analysis and application can be greatly facilitated; all pieces of data of a certain coding object can be displayed in a unified manner, so as to intuitively understand the problems of the coding object, without the need for searching for the pieces of data in the system one by one, thereby improving the data search efficiency; and a panoramic data view with the coding object as the center is formed, thereby providing a good tool and means for quickly analyzing and solving the problems, and improving the degree of structuring of the data.

Classes IPC  ?

• G06F 16/2455 - Exécution des requêtes

Abrégé

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.

Classes IPC  ?

• G06F 11/30 - Surveillance du fonctionnement
• G21D 3/00 - Commande des installations à énergie nucléaire
• G21D 3/04 - Dispositions de sécurité

Abrégé

The present invention relates to the field of pressurized water reactor nuclear power plant coolant boron concentration monitoring. Particularly disclosed are a method and system for improving boron meter measurement accuracy. The method for improving boron meter measurement accuracy in the present invention comprises the following steps: S1, performing temperature compensation for a collected neutron count rate on the basis of a pipeline temperature and a preset temperature compensation coefficient; S2, performing validity determination on the neutron count rate subjected to the temperature compensation; S3, selecting a reference neutron count rate from calibration test data; S4, determining, according to a normalized value of the neutron count rate measured by a boron meter, a measuring range to which the 10B concentration measured by the boron meter belongs; S5, calculating, on the basis of a calibration test fitting coefficient and the normalized value of the neutron count rate, 10B concentrations measured by the boron meter; and S6, calculating and outputting, on the basis of the 10B abundance in a nuclear island primary coolant and a selected and output 10B concentration measured by the boron meter, a total boron concentration measured by the boron meter.

Classes IPC  ?

• G21C 17/022 - Dispositifs ou dispositions pour la surveillance du réfrigérant ou du modérateur pour la surveillance de réfrigérants ou de modérateurs liquides
• G01N 23/00 - Recherche ou analyse des matériaux par l'utilisation de rayonnement [ondes ou particules], p. ex. rayons X ou neutrons, non couvertes par les groupes , ou

Abrégé

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.

Classes IPC  ?

• G06F 30/23 - Optimisation, vérification ou simulation de l’objet conçu utilisant les méthodes des éléments finis [MEF] ou les méthodes à différences finies [MDF]

Abrégé

A hoisting method for a nuclear island main system device. The method comprises the following steps: S1, when a main system device (10) is in a recumbent state, connecting a main crane to the head of the main system device (10), and connecting an auxiliary crane to the tail of the main system device (10); S2, the main crane hoisting the head of the main system device (10), and the auxiliary crane cooperating with the main crane until the main system device (10) changes from the recumbent state to an upright state; and S3, removing the connection between the auxiliary crane and the tail of the main system device (10), and the main crane hoisting the main system device (10) to an installation position. According to the hoisting method for a nuclear island main system device, the hoisting of a main system device(10) is realized by means of the cooperation of two cranes, such that the installation technique and requirements are simple, the installation workload is small, the operation is easy, the efficiency is improved, and the hoisting process of a nuclear island main system device is optimized.

Classes IPC  ?

• B66C 1/12 - Élingues comportant chaînes, fils métalliques, cordes ou bandesFilets
• B66C 13/08 - Dispositifs auxiliaires pour commander les mouvements des charges suspendues ou pour empêcher le câble de prendre du mou pour déposer les charges selon un orientement ou dans une position donnés

Abrégé

The present invention belongs to the technical field of automated nuclear power plant control and relates to a design method for an automatic nuclear power plant start-up and shutdown and intelligent monitoring system. The method comprises the following steps: performing task analysis on a nuclear power plant start-up and shutdown operation process; determining a design range of an automatic nuclear power plant start-up and shutdown system (ANPS); designing function groups of the ANPS; designing a sequence control function flow of the ANPS; designing a main technological parameter closed-loop adjustment function of the ANPS; designing an intelligent monitoring module for the operation state of a unit start-up and shutdown process of the ANPS; performing integrated design on the ANPS; designing a human-machine interface function of the ANPS; and verifying and confirming a design scheme of the ANPS. By means of the method provided in the present invention, the automatic start-up and shutdown sequence control, the main technological parameter closed-loop adjustment, and the intelligent monitoring of a unit operation state parameter of a nuclear power plant in an all-condition range can be realized.

Classes IPC  ?

• G21D 3/04 - Dispositions de sécurité

Abrégé

A design method for a human-machine interface between accident handling guidelines of a nuclear power plant, comprising: a classification step: classifying instructions of unstructured guidelines so as to sort the instructions into corresponding instruction categories; an analysis step: analyzing a parameter requirement of each instruction; and a flowchart design step: designing a guideline flowchart according to a step logical order of the unstructured guidelines, and the categories and required parameter requirements of the instructions. By correlating the unstructured guidelines with each other, the working efficiency of modifying and determining guidelines is improved. By means of a unified and standardized guideline computerization scheme, the efficiency and quality of guideline computerization are improved, and the performance of digital guidelines is ensured.

Classes IPC  ?

• G21D 3/00 - Commande des installations à énergie nucléaire

Abrégé

An electromagnetic detection method for a welding seam defect of a stainless steel pipeline, comprising: acquiring a system noise value a in a gradient signal of a defect-free welding seam area of a reference block, a background noise maximum value b in the gradient signal of the defect-free welding seam area, and a maximum value c of a gradient signal of a defect welding seam position (101); setting an alarm signal value d=the background noise maximum value b*a preset interval, and setting the alarm signal value d

Classes IPC  ?

• G01N 27/82 - Recherche ou analyse des matériaux par l'emploi de moyens électriques, électrochimiques ou magnétiques en recherchant des variables magnétiques pour rechercher la présence des criques

Abrégé

The present invention relates to a quantitative evaluation method for evaluating the independence of defense-in-depth levels for a nuclear power plant, and a system for implementing the method. The method comprises the following steps: (1) determining basic information for quantitative evaluation of the independence of defense-in-depth levels; (2) quantifying part of the basic information according to a probability safety analysis model; and (3) calculating the degrees of relevance between the defense-in-depth levels according to the basic information and quantified quantitative information, so as to obtain a quantitative evaluation result regarding the independence of the defense-in-depth levels. The method and the system provided in the present invention facilitate specific, objective analysis and understanding of the degrees of independence between defense-in-depth levels, compensate for the insufficiency of currently available qualitative techniques, and support and optimize the design of defense-in-depth levels for a nuclear power plant.

Classes IPC  ?

• G06Q 10/06 - Ressources, gestion de tâches, des ressources humaines ou de projetsPlanification d’entreprise ou d’organisationModélisation d’entreprise ou d’organisation

Abrégé

A simplified method for overall criticality safety analysis of a pipeline equipment room, relating to the technical field of criticality safety analysis. The method comprises the following steps: performing, according to a medium in pipelines, screening analysis on all pipelines in an equipment room to be analyzed to screen out pipelines containing a fissile material solution as effective pipelines, and calculating the total solution volume in all effective pipelines (S1); establishing a simplified pipeline bundle model, wherein all the pipelines in the simplified pipeline bundle model are closely arranged according to a minimum arrangement spacing (S2); keeping the total solution volume of the pipeline bundle model unchanged, and by changing the number and height of the pipeline bundle model, searching and calculating the pipeline bundle size for realizing maximum reactivity of the simplified pipeline bundle model (S3); and tightly attaching the simplified pipeline bundle model having the maximum reactivity next to an equipment having the maximum reactivity in the equipment room in a tight arrangement mode for realizing maximum interaction (S4). The method is much simpler than the process of establishing pipelines one by one in a refined manner, improves the calculation and analysis efficiency, and avoids the problem that a calculation result exceeds a criticality safety receiving criterion due to overly conservative and simple gathering of a pipeline solution.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu
• G06F 113/14 - Tuyaux

Abrégé

A method for testing and evaluating the sealing performance of an important residence zone of a nuclear power station, comprising the following steps: step S1, testing the sealing performance of the boundary of an important residence zone; and step S2, testing and evaluating the sealing performance of an area of the important residence zone. The step S2 comprises: step S2.1, arranging a sampling point of a tracer gas in the important residence zone; step S2.2, measuring the free volume of the important residence zone by using a tracer gas method, and evaluating same; step S2.3, testing and evaluating the sealing performance of the area of the important residence zone by using a concentration attenuation method; and step S2.4, testing and evaluating the sealing performance of the area of the important residence zone by using a constant flow injection method. The method covers a complete technical system such as equipment research and development, measuring point arrangement and method application required by testing the sealing performance of the boundary of the important residence zone, and testing and evaluating the sealing performance of the area of the important residence zone, and can be directly used for engineering application.

Classes IPC  ?

• G01M 3/08 - Examen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite par observation de bulles dans un bassin empli de liquide pour tuyaux, câbles ou tubesExamen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite par observation de bulles dans un bassin empli de liquide pour raccords ou étanchéité de tuyauxExamen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite par observation de bulles dans un bassin empli de liquide pour soupapes
• G01M 3/14 - Examen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite par observation d'enveloppes ou de revêtements élastiques, p. ex. eau savonneuse pour tuyaux, câbles ou tubesExamen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite par observation d'enveloppes ou de revêtements élastiques, p. ex. eau savonneuse pour raccords ou étanchéité de tuyauxExamen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite par observation d'enveloppes ou de revêtements élastiques, p. ex. eau savonneuse pour soupapes
• G01M 3/22 - Examen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite en utilisant des révélateurs particuliers, p. ex. teinture, produits fluorescents, produits radioactifs pour tuyaux, câbles ou tubesExamen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite en utilisant des révélateurs particuliers, p. ex. teinture, produits fluorescents, produits radioactifs pour raccords ou étanchéité de tuyauxExamen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite en utilisant des révélateurs particuliers, p. ex. teinture, produits fluorescents, produits radioactifs pour soupapes
• G01M 3/24 - Examen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation d'un fluide ou en faisant le vide par détection de la présence du fluide à l'emplacement de la fuite en utilisant des vibrations infrasonores, sonores ou ultrasonores
• G01N 30/02 - Chromatographie sur colonne

Abrégé

A passive residual heat removal device for a horizontal micro-reactor includes a shielding insulation shell and a deflector. The shielding insulation shell is arranged on an outside of a reactor pressure vessel. The deflector is arranged between the shielding insulation shell and the reactor pressure vessel, to divide a flow space for air in the shielding insulation shell into an inside flow path and an outside flow path. Air in the inside flow path flows from front and rear ends of the deflector to the outside flow path. The shielding insulation shell includes an air inlet and an air outlet, the air inlet connected to the deflector in a sealed manner by an air intake channel. Cold air is introduced into the inside flow path via the air inlet for direct cooling of the reactor pressure vessel. High-temperature air from the outside flow path is discharged via the air outlet.

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur

Abrégé

An online protection method and system. The online protection method comprises: obtaining calibration state data that is periodically updated (S10); performing reconstruction calculation on reactor core power distribution data corresponding to the total rod positions of control rods (S20); obtaining the current total rod position of each control rod in real time, and selecting a group of reactor core power distribution data from a plurality of groups of reactor core power distribution data as the current reactor core power distribution data (S30); calculating a maximum LPD value and a minimum DNBR value corresponding to the current total rod position of the control rod (S40); and according to the calculated maximum LPD value, the calculated minimum DNBR value, a preset LPD setting value, and a preset DNBR setting value, determining whether to trigger a reactor shutdown signal (S50).

Classes IPC  ?

• G21C 17/00 - SurveillanceTests
• G21C 17/108 - Mesure du flux

Abrégé

A passive residual heat removal device and a miniature horizontal reactor system. The device comprises a shielding insulation shell (6) and a deflector (3), wherein the shielding insulation shell (6) is arranged on an outer side of a reactor pressure vessel (7); the deflector (3) is arranged inside the shielding insulation shell (6) and is arranged on the outer side of the reactor pressure vessel (7) in a covering manner to divide a flow space of air in the shielding insulation shell (6) into an inside flow channel and an outside flow channel, with the air in the inside flow channel flowing from front and rear ends of the deflector (3) to the outside flow channel; and the shielding insulation shell (6) is provided with an air inlet (1) and an air outlet (5), the air inlet (1) being connected to the deflector (3) in a sealed manner by means of an air intake channel, cold air from the environment being introduced into the inside flow channel via the air inlet (1) for direct cooling of the reactor pressure vessel (7), and high-temperature air from the outside flow channel being discharged to the environment via the air outlet (5). The optimization for the construction of the flow channel can significantly improve the heat removal capacity without additionally provided auxiliary facilities, thereby meeting the requirements for size limitation and capacity of the residual heat removal device.

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur

Abrégé

An intelligent data reconstruction method and system for a sensor. The method comprises: S1, determining whether a data reconstruction model running condition is met, and if the data reconstruction model running condition is met, entering step S2, otherwise, terminating the steps of data reconstruction; S2, carrying out sensor state detection, determining whether a detection result is abnormal, and if the detection result is shown to be abnormal, entering step S3, and if the detection result is shown to be normal, terminating the steps of data reconstruction; and S3, inputting real-time measurement data of a sensor, which has an abnormal detection result that is obtained in step 2, into a data reconstruction model, so as to obtain a reconstruction value. Whether a data reconstruction model running condition is met and whether a state detection result of a sensor is abnormal are sequentially determined, and real-time measurement data of a sensor, which has an abnormal detection result, is input into a data reconstruction model to obtain a reconstruction value, such that a faulty sensor can be quickly and actively identified after the sensor has a fast fault.

Classes IPC  ?

• G21D 3/00 - Commande des installations à énergie nucléaire
• G21D 3/06 - Dispositions de sécurité réagissant à des défaillances à l'intérieur de l'installation
• G21C 17/00 - SurveillanceTests

Abrégé

A steam-water separator (10) and a waste heat removal system. The steam-water separator (10) comprises a separator body (100) and first connecting pipes (200). A first accommodation cavity (110) is provided in the separator body (100). A water inlet (120) in communication with the first accommodation cavity (110) is provided on one end of the separator body (100). The separator body (100) is provided with a plurality of first connecting holes (130) in communication with the first accommodation cavity (110). A plurality of first connecting pipes (200) are arranged. The number of first connecting pipes (200) is equal to the number of first connecting holes (130). The first connecting pipes (200) are connected to the first connecting holes (130) in a one-to-one correspondence manner.

Classes IPC  ?

• B01D 45/02 - Séparation de particules dispersées dans des gaz ou des vapeurs par gravité, inertie ou force centrifuge par gravité
• G21C 15/16 - Dispositions pour le refroidissement à l'intérieur de l'enceinte sous pression contenant le cœurEmploi de réfrigérants spécifiques comprenant des moyens de séparation du liquide et de la vapeur

Abrégé

An electrical power supply changeover test method, apparatus and device for a nuclear power plant. The method comprises: configuring initial states of a BAS56 test and a BAS54 test in parallel, wherein when the initial state of the BAS54 test is configured, an available state of an auxiliary device for maintaining a hot shutdown working condition of a primary circuit and a vacuum condition of a condenser meets an auxiliary device startup requirement of a power supply for a plant in an auxiliary standby transformer power supply mode (S101); executing the BAS56 test, and after the BAS56 test ends, obtaining a final state of the BAS56 test (S102); in the final state of the BAS56 test, realizing fast power cut-off and transmission between adjacent units of a common load for a plant of a nuclear power unit by using fast power supply changeover technology which simulates automatic changeover (S103); and executing the BAS54 test by taking the final state of the BAS56 test as the initial state of the BAS54 test (S104). By means of the method, the deep decoupling of an electrical power supply changeover test for a plant from a grid-related test can be realized, such that the degree of association between the electrical power supply changeover test for a plant and grid-related work is reduced, thereby realizing the high flexibility of a test window, and further improving the test efficiency, simplifying test operations, and reducing test risks.

Classes IPC  ?

• G21D 3/00 - Commande des installations à énergie nucléaire

Abrégé

The present application relates to the technical field of functional polymer materials, and in particular, to a concrete laitance reinforcement, a preparation method therefor, and an application thereof. The concrete laitance reinforcement provided by the present application is a graft-copolymerized epoxy concrete laitance reinforcement containing a multifunctional group; a main agent is a submicron grafted multifunctional resin having a particle size of 0.05-0.25 μm; a silanol structural unit thereof has excellent permeability to laitance and the inside, and is coupled to surface OH to enhance the strength of a laitance substrate and improve the tensile cohesion strength of the laitance on the surface of concrete; a construction paint also has good strength when the laitance does not need to be polished completely in engineering application; the reinforcement has good wetting adhesion to a concrete substrate and a subsequent coating; and the adhesive force of a matched protective coating is greater than or equal to 1.5 MPa. Moreover, the reinforcement is cross-linked to obtain a thermosetting coating film, such that the heat resistance and temperature creep resistance of the reinforcement are improved, the heat resistance of the reinforcement is improved, and the temperature resistance requirement of the coating at 120°C for 200 h can be met.

Classes IPC  ?

• C09D 183/10 - Copolymères séquencés ou greffés contenant des séquences de polysiloxanes
• C08G 77/442 - Polymères séquencés ou greffés contenant des segments de polysiloxanes contenant des segments de polymères vinyliques
• C04B 41/64 - Composés comportant au moins une liaison carbone-métal ou carbone-silicium

Abrégé

A high-precision measurement method and system for a voltage transformer. The method comprises the following steps: acquiring a system error of a measurement system (S101); performing series resonance on the measurement system in a low-voltage frequency-modulation resonance manner, so as to obtain a resonant frequency point (S102); calculating a resonant inductance according to the resonant frequency point (S103); adjusting a resonant device according to the resonant inductance, so as to boost the power-frequency resonance of the measurement system (S104); when the power-frequency resonance of the measurement system is boosted, acquiring precision data of a voltage transformer under test (S105); and according to the system error and the precision data, measuring the precision of the voltage transformer under test (S106). By means of introducing a system error, the precision of a voltage transformer can be accurately measured in a long-distance and strong-interference environment, such that losses caused by misjudgment are avoided; moreover, check boosting is performed by using a method of first frequency-modulation resonance and then induction-modulation resonance, such that the device weight and volume can be effectively reduced, thereby improving the field check level and the working efficiency.

Classes IPC  ?

• G01R 35/02 - Test ou étalonnage des appareils couverts par les autres groupes de la présente sous-classe des dispositifs auxiliaires, p. ex. des transformateurs pour appareils en fonction du rapport de transformation, de l'angle de phase ou de la puissance à l'utilisation

Abrégé

A homogenized coated particle dispersion fuel and a preparation method therefor. The homogenized coated particle dispersion fuel comprises a matrix material, dressed TRISO coated fuel particles and a series of silicon carbide cylinder bodies (1). The radial uniform distribution of the TRISO coated fuel particles is achieved, the temperature gradient of the coated particle dispersion fuel during operation in a reactor is reduced, the risk of radioactive product release is reduced, and at the same time, the problem of inaccurate neutron physics and thermal hydraulic theory calculation of a gas-cooled microreactor is solved.

Classes IPC  ?

• G21C 3/64 - Combustible céramique en dispersion, p. ex. cermet
• G21C 21/14 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives par placage dans un fluide

Abrégé

The present disclosure relates to a method for analyzing the breakage rate of coated fuel particles taking into consideration manufacturing uncertainty. By means of the method provided in the present disclosure, several groups of coated fuel particles having sizes that satisfy a manufacturing uncertainty distribution can be constructed by means of a statistical sampling method; the several groups of coated fuel particles are then analyzed in combination with currently mature analysis program software for the breakage rate of coated fuel particles, so as to obtain several corresponding groups of analysis results of the breakage rates of the coated fuel particles; the analysis results of the breakage rates of the coated fuel particles are then processed by using a statistical analysis method; and finally, the breakage rate of the coated fuel particles that meets a confidence interval requirement and takes manufacturing uncertainty into consideration is obtained.

Classes IPC  ?

• G06F 17/18 - Opérations mathématiques complexes pour l'évaluation de données statistiques

Abrégé

An online monitoring method for wall surface abrasion of a concrete volute pump, comprising: acquiring an actual hydraulic head H' and a preset warning roughness of the concrete volute pump under the working condition of flow Q, calculating a warning hydraulic head at the position of the flow Q according to a flow-hydraulic head-roughness relational expression, or converting the preset warning roughness into a roughness coefficient, calculating a warning hydraulic head at the position of the flow Q according to a flow-hydraulic head-roughness coefficient relational expression, and when the actual hydraulic head H' is lower than the warning hydraulic head, giving an alarm for the abrasion condition of a concrete volute.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

Abrégé

A homogenized coated particle dispersion fuel and a preparation method therefor. The homogenized coated particle dispersion fuel comprises a matrix material, dressed TRISO coated fuel particles and a series of silicon carbide cylinder bodies (1). The radial uniform distribution of the TRISO coated fuel particles is achieved, the temperature gradient of the coated particle dispersion fuel during operation in a reactor is reduced, the risk of radioactive product release is reduced, and at the same time, the problem of inaccurate neutron physics and thermal hydraulic theory calculation of a gas-cooled microreactor is solved.

Classes IPC  ?

• G21C 3/64 - Combustible céramique en dispersion, p. ex. cermet
• G21C 21/14 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives par placage dans un fluide

Abrégé

Provided in the present invention are an experimental research system and method for retention of an aerosol in a containment in a micro-channel. An aerosol delivery system quantitatively conveys an aerosol into a simulation container; a steam supply system quantitatively conveys steam into the simulation container; the simulation container, a micro-channel structure, and an aerosol concentration measurement system are connected in sequence; the simulation container mixes the aerosol and the steam, and maintains a certain temperature and pressure to simulate an in-containment environment under a severe accident; the micro-channel structure is configured to simulate a containment gap environment; mixed gas formed by mixing the aerosol and the steam is retained by means of the micro-channel structure and then leaks into the aerosol concentration measurement system; and the aerosol concentration measurement system is configured to measure the aerosol concentration in the mixed gas. According to the present invention, by means of the design of a brand-new experimental research system for retention of an aerosol in a containment in a micro-channel, research on retention characteristics of the aerosol on the containment in a small gap under severe accident working conditions can be reproduced, and a basis is provided for research on a retention characteristic model.

Classes IPC  ?

• G01N 15/06 - Recherche de la concentration des suspensions de particules

Abrégé

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.

Classes IPC  ?

• F28C 1/00 - Réfrigérants à ruissellement à contact direct, p. ex. tours de réfrigération
• F28F 25/10 - Parties constitutives des réfrigérateurs à ruissellement pour l'alimentation en gaz ou en vapeur
• F28F 19/01 - Prévention de la formation de dépôts ou de la corrosion, p. ex. en utilisant des filtres en utilisant des moyens pour séparer les éléments solides du fluide échangeur de chaleur, p. ex. des filtres
• F28F 25/02 - Parties constitutives des réfrigérateurs à ruissellement pour répartir, faire circuler ou collecter le liquide
• F28F 25/08 - Planches ou grilles d'éclaboussement, p. ex. pour convertir la pulvérisation de liquides en films de liquidesÉléments ou couches pour accroître l'aire de la surface de contact
• F28F 25/06 - Buses ou canalisations de pulvérisation
• F28F 25/04 - Augets répartiteurs ou collecteurs
• G21D 1/02 - Dispositions d'équipement auxiliaire

Abrégé

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.

Classes IPC  ?

• F22B 33/18 - Combinaisons de chaudières à vapeur avec d'autres appareils
• F22D 11/02 - Aménagement des pompes d'alimentation en eau
• F24D 1/08 - Distribution par conduits, p. ex. à conduit unique

Abrégé

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.

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur
• G21D 3/06 - Dispositions de sécurité réagissant à des défaillances à l'intérieur de l'installation

Abrégé

Disclosed in the present invention are a vertical spring, a grid and a fuel assembly. The vertical spring is provided on a strap of the grid and is located in a lattice cell of the grid, and the vertical spring comprises a strip, which protrudes towards the center direction of the lattice cell, and at least one flow guide hole, which is arranged in the strip; and the flow guide hole penetrates two opposite surfaces of the strip. In the present invention, the vertical spring is arranged on the strap of the grid and is located in the lattice cell of the grid, the flow guide hole is arranged in the strip of the vertical spring, and a fluid, which flows to a triangular area between a fuel rod and the strip, is guided by means of the flow guide hole, such that the fluid is prevented from being collected near a point wherein the strip and the fuel rod come into contact, transverse flow disturbance caused by the vertical spring is reduced, and the problem of excessive pressure drop caused by a grid spring in the prior art is solved.

Classes IPC  ?

• G21C 3/34 - Grilles d'espacement
• G21C 3/356 - Grilles d'espacement munies d'organes de support des éléments combustibles

Abrégé

Disclosed in the present invention are a positioning lattice and a fuel assembly. The positioning lattice is used for the middle position of the fuel assembly, and is made of a material having the same or greater expansion coefficient with respect to a reactor core plate material. The positioning lattice of the present invention is made of the material having the same or greater expansion coefficient with respect to the reactor core plate material, is mainly used for the middle position of the fuel assembly, and effectively reduces a gap between lattices under normal operation conditions, thereby improving the safety performance of a reactor, and reducing the uncertainty of thermal work caused by the gap at the same time.

Classes IPC  ?

• G21C 3/30 - Assemblages d'un certain nombre d'éléments combustibles sous forme d'une unité rigide
• G21C 3/12 - Moyens formant partie de l'élément pour le positionner dans le cœur du réacteurEntretoises extérieures à cet effet

Abrégé

A heat exchange water tank liquid level measurement system and method, and a nuclear power plant PCS. A first pressure measurement instrument of the liquid level measurement system is located on a water outlet pipeline (5) at the bottom of a heat exchange water tank (8), and is used for measuring a water pressure at the bottom of the heat exchange water tank (8); a second pressure measurement instrument of the liquid level measurement system is located on a manhole channel (7) at the top of the heat exchange water tank (8), and is used for measuring a water pressure at the top of the heat exchange water tank (8); a DCS system (1) of the liquid level measurement system is used for calculating a pressure difference, and according to the logical formula of the liquid level and pressure difference stored therein, calculating the liquid level in the heat exchange water tank (8). Thus, the problem of poor instrument precision caused by long-distance capillary transmission is avoided, and the risk of installation and the difficulty of maintenance and measurement are greatly reduced.

Classes IPC  ?

• G21C 17/035 - Dispositifs de détection du niveau de réfrigérant ou du modérateur
• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur
• G01F 23/18 - Indication ou mesure du niveau des liquides ou des matériaux solides fluents, p. ex. indication en fonction du volume ou indication au moyen d'un signal d'alarme par mesurage de la pression les dispositifs d'indication, d'enregistrement ou d'alarme étant actionnés électriquement

Abrégé

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.

Classes IPC  ?

• F16K 31/122 - Moyens de fonctionnementDispositifs de retour à la position de repos actionnés par un fluide le fluide agissant sur un piston
• F15B 11/08 - Systèmes de servomoteurs dépourvus d'asservissement avec un seul servomoteur

Abrégé

A threaded member (10) manufacturing method and a thread machining device. The threaded member (10) manufacturing method comprises: S1, removing a failed screw section on a threaded member (10) and polishing an exposed cross section (110) flat; wherein a failure includes at least one of breakage, corrosion, and thread damage; S2, sleeving and fixing a connecting sleeve (20) on the periphery of an embedded section (11) of the threaded member (10), and polishing the top surface of the connecting sleeve (20) to be flush with the cross section (110); S3, forming a screw blank section (30) on the embedded section (11) and the connecting sleeve (20) by means of laser cladding; S4, machining the screw blank section (30) to a polished rod that is coaxial with the embedded section (11); S5, machining threads on the polished rod to form a new screw section (12), and connecting to the embedded section (11) to form an integral threaded member (10).

Classes IPC  ?

• B23G 1/00 - FiletageMachines automatiques spécialement conçues pour le filetage
• B23G 1/44 - Équipement ou accessoires spécialement conçus pour les machines ou dispositifs à fileter

Abrégé

The present invention relates to a method for diagnosing the performance of a control valve of a steam turbine of a nuclear power station, and a storage medium and a computer. The method comprises: S1, establishing a hydraulic servo-motor resistance manufacturer database and a hydraulic servo-motor resistance installation database; S2, establishing an intelligent diagnosis system to perform resistance and performance measurement on a steam intake valve; S3, performing resistance measurement of the steam intake valve before steam is introduced into the steam intake valve; S4, performing resistance measurement of the steam intake valve after steam is introduced into the steam intake valve; and S5, the intelligent diagnosis system performing, on the steam intake valve and according to obtained test data, intelligent diagnosis that combines qualitative analysis and quantitative analysis. By means of the present invention, an online fault diagnosis basis for diagnosing faults, such as a quick closing time exceeding the standard and jamming of a main steam valve and a regulating valve of a steam turbine, is provided, and an online measurement method for resistance measurement of a hydraulic servo-motor is provided, such that the maintenance time can be shortened, thereby reducing the number of false stops of a unit, and improving the safety and economy of the unit.

Classes IPC  ?

• G06Q 10/06 - Ressources, gestion de tâches, des ressources humaines ou de projetsPlanification d’entreprise ou d’organisationModélisation d’entreprise ou d’organisation

Abrégé

A probabilistic safety analysis-based fire protection selection method and apparatus for cables in a nuclear power plant. The method comprises: performing quantitative analysis on an improved fire probabilistic safety analysis model, and calculating and obtaining a fire risk result of the nuclear power plant; obtaining an initial fire protection coated cable list, recognizing and modeling cables in the initial fire protection coated cable list in the model, and obtaining a fire risk quantification result; and determining whether the fire risk quantification result meets a probabilistic safety objective, and if not, adding other cables affected by fire to the initial fire protection coated cable list until the fire risk quantification result meets the probabilistic safety objective, so as to obtain a cable fire protection target list. By means of the method, cables having critical risks can be quantitatively recognized as cables that require fire protection, and the fire protection investment of the nuclear power plant is optimized, such that the maximum balance between safety and economy is achieved.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

Abrégé

A system for testing a containment building of a nuclear power plant, the system comprising: a containment building overall leakage rate measurement module (10), a containment building acoustic leakage detection module (20), an alarm module (100), and a display module (90). The containment building overall leakage rate measurement module (10) is used to measure the overall leakage rate in a containment building and perform calculation on measured data to obtain the real-time overall leakage rate of the containment building and the degree of uncertainty; the containment building acoustic leakage detection module (20) is used to monitor sound signals of the containment building and output containment building acoustic leakage detection measurement results after analyzing and processing the monitored sound signals; the alarm module (100) outputs a corresponding alarm signal in the case of the real-time overall leakage rate, the degree of uncertainty, and containment building acoustic leakage; and the display module (90) displays the measurement results of the real-time overall leakage rate, the degree of uncertainty, and the containment building acoustic leakage.

Classes IPC  ?

• G21C 17/003 - Inspection à distance des enceintes, p. ex. des cuves de pression
• G01N 29/04 - Analyse de solides
• G01N 29/14 - Recherche ou analyse des matériaux par l'emploi d'ondes ultrasonores, sonores ou infrasonoresVisualisation de l'intérieur d'objets par transmission d'ondes ultrasonores ou sonores à travers l'objet utilisant des techniques d'émission acoustique

Abrégé

A silver-loaded zeolite filter, and a nuclear power plant containment filtering and discharging system comprising the silver-loaded zeolite filter. The silver-loaded zeolite filter comprises a pressure-bearing tank (1), a silver-loaded zeolite filtering unit (3) and a gas collection header (4), wherein an inlet is provided in the top of the pressure-bearing tank (1), an outlet is provided in the bottom of the pressure-bearing tank (1), gas to be filtered enters the pressure-bearing tank (1) through the inlet, the silver-loaded zeolite filtering unit (3) is arranged in the pressure-bearing tank (1) and is used for removing organic iodine in the gas to be filtered, the gas collection header (4) is arranged below the silver-loaded zeolite filtering unit (3) and is in communication with the bottom end of the silver-loaded zeolite filtering unit (3) and used for collecting the filtered gas, and the outlet of the pressure-bearing tank (1) is in communication with the bottom end of the gas collection header (4).

Classes IPC  ?

• B01D 46/00 - Filtres ou procédés spécialement modifiés pour la séparation de particules dispersées dans des gaz ou des vapeurs
• G21C 13/10 - Moyens pour prévenir la contamination dans le cas d'une fuite

Abrégé

A gas-cooled micro-reactor core, comprising a reflecting layer, fuel units, and a control rod assembly (2), wherein the control rod assembly (2) and the fuel units are all arranged in the reflecting layer, a plurality of fuel units are provided, each fuel unit comprises a pressure pipe (8) and a fuel assembly (1), the fuel assembly (1) is arranged in the pressure pipe (8) so as to make the pressure pipe (8) act as a pressure bearing boundary of the reactor core, and the control rod assembly (2) is arranged outside the pressure pipe (8). The reactor core is easy to mount, and solves the problems of many elements to be transported, a complex assembly, and a slow response to deployment of a high-temperature gas-cooled reactor.

Classes IPC  ?

• G21C 3/326 - Faisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant des éléments combustibles de différentes compositionsFaisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant, en plus des éléments combustibles, d'autres éléments en forme d'aiguille, de barre ou de tube, p. ex. barres de commande, barres de support de grilles, barres fertiles, barres à poison ou barres factices
• G21C 5/02 - Structure du modérateur ou du cœurEmploi de matériaux spécifiés comme modérateur Détails

Abrégé

A reactor core system (6), comprising a reactor core, a reflective layer and a rotary drum control rod (2), wherein the reactor core, which is transversely arranged, comprises a plurality of fuel assemblies (1), and the plurality of fuel assemblies (1) are arranged according to a radial partition and in an axial layered manner; the reflective layer wraps outside the reactor core; the rotary drum control rod (2) is arranged in the reflective layer; a central graphite belt (5) is arranged in the reactor core, and the central graphite belt (5) is arranged in a vertical direction; an absorption body ball channel (4) is arranged in the central graphite belt (5), and the absorption body ball channel (4) is vertically arranged and penetrates the reflective layer; and an absorption body ball is arranged in the absorption body ball channel (4). The reactor core system (6) is small in size and convenient to transport.

Classes IPC  ?

• G21C 3/326 - Faisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant des éléments combustibles de différentes compositionsFaisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant, en plus des éléments combustibles, d'autres éléments en forme d'aiguille, de barre ou de tube, p. ex. barres de commande, barres de support de grilles, barres fertiles, barres à poison ou barres factices
• G21C 7/08 - Commande de la réaction nucléaire par application de matériau absorbant les neutrons, c.-à-d. matériau avec section efficace d'absorption excédant largement la section efficace de réflexion par déplacement des éléments de commande solides, p. ex. barres de commandes

Abrégé

A reactor core system (6), comprising a reactor core, a reflective layer and a rotary drum control rod (2), wherein the reactor core, which is transversely arranged, comprises a plurality of fuel assemblies (1), and the plurality of fuel assemblies (1) are arranged according to a radial partition and in an axial layered manner; the reflective layer wraps outside the reactor core; the rotary drum control rod (2) is arranged in the reflective layer; a central graphite belt (5) is arranged in the reactor core, and the central graphite belt (5) is arranged in a vertical direction; an absorption body ball channel (4) is arranged in the central graphite belt (5), and the absorption body ball channel (4) is vertically arranged and penetrates the reflective layer; and an absorption body ball is arranged in the absorption body ball channel (4). The reactor core system (6) is small in size and convenient to transport.

Classes IPC  ?

• G21C 7/08 - Commande de la réaction nucléaire par application de matériau absorbant les neutrons, c.-à-d. matériau avec section efficace d'absorption excédant largement la section efficace de réflexion par déplacement des éléments de commande solides, p. ex. barres de commandes
• G21C 3/326 - Faisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant des éléments combustibles de différentes compositionsFaisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant, en plus des éléments combustibles, d'autres éléments en forme d'aiguille, de barre ou de tube, p. ex. barres de commande, barres de support de grilles, barres fertiles, barres à poison ou barres factices

Abrégé

A gas-cooled micro-reactor core, comprising a reflecting layer, fuel units, and a control rod assembly (2), wherein the control rod assembly (2) and the fuel units are all arranged in the reflecting layer, a plurality of fuel units are provided, each fuel unit comprises a pressure pipe (8) and a fuel assembly (1), the fuel assembly (1) is arranged in the pressure pipe (8) so as to make the pressure pipe (8) act as a pressure bearing boundary of the reactor core, and the control rod assembly (2) is arranged outside the pressure pipe (8). The reactor core is easy to mount, and solves the problems of many elements to be transported, a complex assembly, and a slow response to deployment of a high-temperature gas-cooled reactor.

Classes IPC  ?

• G21C 3/326 - Faisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant des éléments combustibles de différentes compositionsFaisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant, en plus des éléments combustibles, d'autres éléments en forme d'aiguille, de barre ou de tube, p. ex. barres de commande, barres de support de grilles, barres fertiles, barres à poison ou barres factices
• G21C 5/02 - Structure du modérateur ou du cœurEmploi de matériaux spécifiés comme modérateur Détails