Abrégé

A portable internal friction measuring instrument and internal friction measurement method, the portable internal friction measuring instrument comprising: a cavity (10), a sample fixture (20), a fixing plate (30), a heating plate (40), a thermal barrier cover (50), an excitation assembly (60), and a vibration sensor (70). The excitation assembly (60) comprises a rotatable vibration impeller (61), at least one protrusion is provided on the outer wall of the vibration impeller (61), the lower edge of a sample (80) can be intermittently toggled by means of the protrusion during the rotation of the vibration impeller (61), and the vibration amplitude of the sample (80) is determined by using an angle difference between emitted light and reflected light of the sample (80), such that the technical effects of mechanical toggling to excite the vibration of the sample (80) and non-contact measurement of the vibration of the sample (80) are achieved.

Classes IPC  ?

• G01N 19/02 - Mesure du coefficient de frottement entre matériaux

Abrégé

A heat and mass transfer analysis method and apparatus for a corrosion product in a nuclear reactor, a computer device, a storage medium, and a computer program product. The heat and mass transfer analysis method for a corrosion product in a nuclear reactor comprises: acquiring the average thermal conductivity of a corrosion product, and on the basis of the average thermal conductivity, determining the temperature distributions of the corrosion product when a coolant is in a saturated state, a nucleate boiling state, and an evaporated state, respectively (S202); on the basis of the pressure distribution in the corrosion product, determining the flow velocity distribution of the coolant in the corrosion product (S204); acquiring a concentration relational expression between the concentration of boric acid and the concentration of another soluble substance in the coolant, and on the basis of the concentration relational expression and the concentration of boric acid, determining the concentration distribution of the another soluble substance (S206); and on the basis of the temperature distributions, the flow velocity distribution, and the concentration distribution of the soluble substance, analyzing heat and mass transfer inside the corrosion product to obtain the concentration of boron after enrichment inside the corrosion product, and on the basis of the concentration after enrichment, obtaining the total amount of boron deposition (S208).

Classes IPC  ?

• G21C 17/00 - SurveillanceTests

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

The present application relates to a reactor power control method and apparatus, a computer device and a readable storage medium. The method comprises: acquiring an initial load signal sent by a power grid; preprocessing the initial load signal to obtain an actual load signal; acquiring first feedforward information on the basis of the actual load signal and a preset load-rod position model; acquiring second feedforward information on the basis of the initial load signal, the actual load signal and a preset deviation intensity adjustment model; acquiring third feedforward information on the basis of the actual load signal and a preset amplitude limiting model; and generating a power control signal on the basis of the first feedforward information, the second feedforward information, the third feedforward information and an actual rod position value of a control rod of a reactor.

Classes IPC  ?

• G21D 3/08 - Régulation de différents paramètres dans l'installation
• H02J 3/24 - Dispositions pour empêcher ou réduire les oscillations de puissance dans les réseaux
• G21D 3/00 - Commande des installations à énergie nucléaire
• G21D 3/10 - Régulation de différents paramètres dans l'installation par une combinaison d'une variable dérivée du flux de neutrons avec d'autres variables de commande, p. ex. dérivées de la température, du flux du réfrigérant, de la pression
• G05B 19/04 - Commande à programme autre que la commande numérique, c.-à-d. dans des automatismes à séquence ou dans des automates à logique

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

A smart cleaning vehicle for solar thermal power station reflecting mirror surfaces and heat collecting tubes, comprising: a vehicle body (7), comprising a driver's compartment (71), a chassis (72), a compressed air system (73), a hydraulic system (74), a fuel steam generator (75), and a first driving system; a mechanical arm (8), mounted on the chassis (72); a cleaning tool frame (0), mounted on the mechanical arm (8) and capable of pitching and rotating along with the mechanical arm (8); and a water path system (9), mounted on the chassis (72) and used for supplying water to the cleaning tool frame (0), wherein the water path system (9) comprises a cleaning pipeline (91) and a steam pipeline (92), the cleaning pipeline (91) is in communication with the cleaning tool frame (0), the steam pipeline (92) is in communication with the fuel steam generator (75) and is then in communication with the cleaning tool frame (0), the compressed air system (73) provides power for the cleaning tool frame (0) and provides purging power for the water path system (9).

Classes IPC  ?

• B60P 3/30 - Véhicules d'arrosage
• B08B 1/00 - Nettoyage par des procédés impliquant l'utilisation d'outils
• B08B 1/04 - Nettoyage par des procédés impliquant l'utilisation d'outils, de brosses ou d'éléments analogues utilisant des éléments actifs rotatifs
• B08B 3/02 - Nettoyage par la force de jets ou de pulvérisations
• B60K 25/02 - Entraînements auxiliaires directs à partir de l'arbre du moteur

Abrégé

An analysis method and apparatus for measurement uncertainty of power distribution of a reactor core, a device, a storage medium, and a product. The analysis method for measurement uncertainty of power distribution of a reactor core comprises: constructing a simulated operation state on the basis of an actual operation state of a reactor core, determining state parameters of the reactor core in the simulated operation state, and acquiring a group constant of the reactor core (S202); performing calculation on the basis of a material calculation subprogram and the state parameters to obtain material components of each fuel rod in the reactor core (S204); performing calculation on the basis of a power calculation subprogram and the material components of each fuel rod to obtain real power distribution of the reactor core, and acquiring detector reaction intensity of the reactor core (S206); on the basis of the state parameters and the group constant, determining power parameters of the reactor core in the simulated operation state, and performing flux graph simulation on the basis of the power parameters and the detector reaction intensity to obtain simulated power distribution of the reactor core (S208); and determining the measurement uncertainty of power distribution of the reactor core on the basis of a difference between the real power distribution and the simulated power distribution (S210).

Classes IPC  ?

• G21C 17/00 - SurveillanceTests
• G21C 17/108 - Mesure du flux
• G21C 17/06 - Dispositifs ou dispositions pour la surveillance ou le test du combustible ou des éléments combustibles en dehors du cœur du réacteur, p. ex. pour la consommation ou pour la contamination
• G06F 17/18 - Opérations mathématiques complexes pour l'évaluation de données statistiques

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

The present application relates to a method and an apparatus for depicting a grid model of a nuclear power wire-wrapped rod bundle, and a device. The method comprises: constructing an initial grid model for a target wire-wrapped rod bundle assembly; extracting at least two control points from a boundary surface of the initial grid model of the target wire-wrapped rod bundle assembly; mapping each control point in the initial grid model of the target wire-wrapped rod bundle assembly to the target wire-wrapped rod bundle assembly to obtain a target mapping point of each control point; and, according to the initial grid model of the target wire-wrapped rod bundle assembly and a position variation between each control point and the corresponding target mapping point of the control point, determining a wire-wrapped rod bundle grid model for the target wire-wrapped rod bundle assembly. The method can improve the precision of depicting the wire-wrapped rod bundle grid model, thereby enabling accurate prediction of a three-dimensional flow field of a coolant in a nuclear power wire-wrapped assembly.

Classes IPC  ?

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

Abrégé

A nuclear reactor fault diagnosis method and apparatus, a computer device, and a storage medium. The method comprises: acquiring system simulation operation data of a nuclear reactor under different candidate working conditions (S201); according to the system simulation operation data, determining a norm value corresponding to each type of fault working condition (S202); if a situation where the norm value is less than a preset norm threshold is present in each type of fault working condition, performing feature engineering processing on the system simulation operation data, and on the basis of the processed system simulation operation data, redetermining the norm value corresponding to each type of fault working condition until the norm value corresponding to each type of fault working condition is greater than or equal to the preset norm threshold (S203); and according to the processed system simulation operation data and system current-operation data of the nuclear reactor under the current working condition, performing fault diagnosis on the current working condition of the nuclear reactor (S204). By means of said method, the sensitivity and accuracy of fault diagnosis of a nuclear reactor can be improved.

Classes IPC  ?

• G06F 18/2433 - Perspective d'une seule classe, p. ex. une classification "une contre toutes"Détection de nouveautéDétection de valeurs aberrantes

Abrégé

A nuclear power plant operation protection method and apparatus, and a computer device and a storage medium. The nuclear power plant operation protection method comprises: if it is detected that there is a failed neutron detector in a nuclear power plant, outputting a power increase block signal to a nuclear power device of the nuclear power plant, such that the nuclear power device operates on the basis of the block signal (S201); according to an identity identifier of the failed neutron detector, calling, from an inverse matrix library, an inverse matrix corresponding to the failed neutron detector (S202); updating three-dimensional power of a reactor core according to the inverse matrix (S203); and according to the updated three-dimensional power of the reactor core, releasing the power increase block signal of the nuclear power device of the nuclear power plant, so as to recover the normal operation of the nuclear power device of the nuclear power plant (S204).

Classes IPC  ?

• G21D 3/06 - Dispositions de sécurité réagissant à des défaillances à l'intérieur de l'installation
• G06F 18/24 - Techniques de classification

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

A nuclear power supply reactor core structure and a processing method therefor. The nuclear power supply reactor core structure comprises a reactor core base body (1), fuel channels (2), and gas channels (3); the reactor core base body (1) is made of metal; the fuel channels (2) are arranged in the reactor core base body (1); the gas channels (3) are used for circulation of a cooling medium, a plurality of gas channels (3) are provided, and the gas channels (3) are circumferentially arranged on the outer sides of the fuel channels (2). The processing method for the nuclear power supply reactor core structure comprises: obtaining a plurality of sheet layers; performing hole processing on the sheet layers in the axial direction of the sheet layers to form fuel channels (2) and gas channels (3); and stacking and fixing adjacent sheet layers in the axial direction of the sheet layers.

Classes IPC  ?

• G21C 5/14 - Structure du modérateur ou du cœurEmploi de matériaux spécifiés comme modérateur caractérisée par la forme

Abrégé

A method and apparatus for determining a transient neutron fluence rate of a core, and a device and a storage medium. The method comprises: when a core is in an unstable state, determining an amplitude value of the core at the previous moment according to a neutron fluence rate of the core at the previous moment (S101); determining a shape value and amplitude value of the core at the current moment according to the amplitude value of the core at the previous moment (S102); and determining a target neutron fluence rate of the core at the current moment according to the shape value and amplitude value of the core at the current moment (S103).

Classes IPC  ?

• G21C 17/108 - Mesure du flux

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

A stress corrosion test device (1) for a grid spring (10), comprising an autoclave (2), a loader (3) and, arranged in the autoclave (2), an upper jig assembly and a lower jig assembly; the loader (3) is used for driving the upper jig assembly to move; the upper jig assembly and the lower jig assembly are respectively used for fixing a fuel cladding (9) and a grid spring (10); the lower jig assembly comprises a support plate (71) fixed to the bottom portion of the autoclave (2) and a pressing plate (72) fitting the support plate (71); the upper jig assembly comprises a clamping plate (61) and a jig plate (62), the clamping plate (61) penetrating through the fuel cladding (9) and then being fixed to the jig plate (62). The stress corrosion test device (1) for a grid spring (10) can effectively implement fixation and apply a loading condition conforming to the actual service environment of the grid spring (10) so as to simulate the actual service environment, and simulate the service state of the grid spring (10) in terms of three main stress corrosion influence factors, i.e. the material, the service environment and the stress state, thus more realistically evaluating the stress corrosion sensitivity of the grid spring (10).

Classes IPC  ?

• G01N 3/08 - Recherche des propriétés mécaniques des matériaux solides par application d'une contrainte mécanique par application d'efforts permanents de traction ou de compression
• G01N 3/18 - Exécution de tests à des températures élevées ou basses
• G01N 3/04 - Mandrins
• G01N 3/02 - Recherche des propriétés mécaniques des matériaux solides par application d'une contrainte mécanique Parties constitutives

Abrégé

A driving device for a curved top cover, a positioning checker, and a positioning and detecting method. The driving device (1) comprises: a driving mechanism (11), comprising a driving frame (111), and a driving assembly (112) mounted on the driving frame (111), wherein the bottom of the driving frame (111) is provided with a wedge assembly attached to a top surface (01); and a flexible guide mechanism (12), which is configured to move the driving mechanism (11), which is on the top surface (01), along a check channel, and comprises a winding frame (121), and a chain (122) wound on the winding frame (121), wherein a free end of the chain (122) is fixedly connected to a front end of the driving mechanism (11). By virtue of a vertical arrangement of a plane on which a motion track is located when a laser emitter (21) and the chain (122) extend out, position calibration of a probe (31) in a first channel and a second channel can be rapidly achieved, and in combination with a characteristic that a bending direction of the chain (122) is limited, a projection of a moving track of the driving mechanism (11) on a bottom surface of the top cover (0) extends along a straight line to ensure movement along the check channel, thereby ensuring that the probe (31) can be attached to the top cover (0) while moving along the check channel.

Classes IPC  ?

• G21C 17/003 - Inspection à distance des enceintes, p. ex. des cuves de pression
• G01N 33/00 - Recherche ou analyse des matériaux par des méthodes spécifiques non couvertes par les groupes

Abrégé

A constant protective value processing method for a nuclear instrument system, and a system. In the constant protective value processing method for a nuclear instrument system, when a reactor core unit operates at a power above 100% FP, a low constant protective value of an intermediate range is in a locked state, and the low constant protective value corresponds to a power below 100% FP of the reactor core unit. The method comprises: a protection step: S11, involving: acquiring the current value of each intermediate-range channel; S12, determining whether the number of intermediate-range channels which have current values exceeding a high constant protective value meets a preset condition, wherein the high constant protective value corresponds to a power above 100% FP, and each intermediate-range channel corresponds to one high constant protective value; and S13, if the number of intermediate-range channels which have current values exceeding the high constant protective value meets the preset condition, triggering a reactor shutdown operation. Power-exceeding reactor shutdown protection is performed by using a high-power constant protective value of an intermediate range, such that a power-exceeding risk of a reactor can be effectively prevented, and a safety risk is prevented from occurring in the reactor when a power range fails, thereby ensuring the safety of the reactor.

Classes IPC  ?

• G21C 17/108 - Mesure du flux

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 relates to a method and apparatus for quantifying indication signals of a nuclear instrument system, and a device, and a storage medium. The method comprises: simulating the power distribution of a reactor core under various operating conditions, so as to obtain power distribution parameters of the reactor core under the various operating conditions; and according to the power distribution parameters of the reactor core under the different operating conditions and according to corresponding benchmark power distribution parameters, quantifying indication signals of the reactor core. The use of the present method allows for quantifying the uncertainty of indication signals of the reactor core by means of theoretical analysis, allowing for obtaining more-accurate indication signals; moreover, the present solution does not require the collection of actual measurement data under actual operating conditions, and is suited to controlling novel reactor cores lacking actual measurement data and in designing constant values of protection signals, and therefore the present solution is highly applicable.

Classes IPC  ?

• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu
• G21C 17/00 - SurveillanceTests

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 application relates to a steam generator water-level control method and system after shutdown of a pressurized water reactor. The method comprises: according to a thermal power load signal of a reactor core during reactor shutdown, generating a steam generator water-level setting value and a steam generator water-supply flow signal under the reactor shutdown; generating a reactor shutdown steam generator water-level control signal; executing steam generator water-level control logic; and switching the steam generator water-level control logic into automatic control after a preset duration. The problem of water-level deviation, which is caused by means of thermal hydraulic parameter perturbation after reactor shutdown in different initial states, of a steam generator being too large is solved, and a reactor device is prevented from being damaged because the water level of the steam generator is too high or too low, thereby preventing abnormal adjustment of the steam generator after reactor shutdown, and thus improving the stability of the steam generator.

Classes IPC  ?

• F22B 35/00 - Systèmes de commande pour chaudières à vapeur
• F22B 37/46 - Utilisation, agencement ou disposition des dispositifs d'alarme ou des dispositifs automatiques de sécurité sensibles à un niveau d'eau trop bas ou trop élevé, p. ex. limitant, interrompant ou éteignant la combustion dans les chaudières
• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur

Abrégé

A method for analyzing a characteristic of a molten pool, which method belongs to the technical field of computers. The method comprises: acquiring a natural convection intensity value of a molten pool in a static state, and a forced convection intensity value of the molten pool in a swing state (S201); then, determining a convection influence factor of the molten pool according to the natural convection intensity value and the forced convection intensity value (S202); finally, according to a relationship between the convection influence factor and a preset threshold value, determining whether there is forced convection in the swing state inside the molten pool (S203); and if there is the forced convection, determining a convection characteristic of the molten pool in the swing state according to the height and transverse span of a free liquid level in the molten pool, and an acceleration in the horizontal direction and an acceleration in the vertical direction of the molten pool in the swing state (S204). The problem of there being currently a lack of effective methods for analyzing a characteristic of a molten pool, especially a lack of methods capable of determining a characteristic of a molten pool in a swing state is solved. Further provided are an apparatus for analyzing a characteristic of a molten pool, and a computer device and a storage medium.

Classes IPC  ?

• G01N 25/16 - Recherche ou analyse des matériaux par l'utilisation de moyens thermiques en recherchant le coefficient de dilatation thermique
• G01N 25/18 - Recherche ou analyse des matériaux par l'utilisation de moyens thermiques en recherchant la conductivité thermique
• G21C 15/00 - Dispositions pour le refroidissement à l'intérieur de l'enceinte sous pression contenant le cœurEmploi de réfrigérants spécifiques

Abrégé

The present invention relates to a method for constructing a risk monitoring and early warning tool of a fuel cell vehicle-mounted hydrogen system. The method comprises the following steps: constructing a device database of a fuel cell vehicle-mounted hydrogen system and a hydrogen leakage database; processing the device database by using a minimum cutset algorithm, so as to obtain a first risk evaluation result, and performing physical analysis on the hydrogen leakage database to obtain a second risk evaluation result; and comprehensively processing the first risk evaluation result and the second risk evaluation result and and performing display. The present invention can implement monitoring and early warning on a risk of a vehicle-mounted hydrogen system, provides guidance for safety management of a hydrogen fuel cell vehicle, and can promote the development and widespread application of hydrogen fuel cell technology.

Classes IPC  ?

• H01M 8/04298 - Procédés de commande des éléments à combustible ou des systèmes d’éléments à combustible
• G06Q 10/06 - Ressources, gestion de tâches, des ressources humaines ou de projetsPlanification d’entreprise ou d’organisationModélisation d’entreprise ou d’organisation

Abrégé

The present application relates to a nuclear reactor optimization method and apparatus, a computer device and a storage medium. The method comprises: according to an optimization requirement with regard to a nuclear reactor, constructing a first objective function; according to the first objective function and the function relationship between an intermediate characteristic variable and an optimization parameter, constructing a second objective function comprising the intermediate characteristic variable; according to a candidate design solution for the nuclear reactor, performing probabilistic modeling on the second objective function to obtain probability distribution of the second objective function; and, according to the probability distribution and the optimization requirement, determining a target value of the optimization parameter. In the present method, the introduction of the intermediate characteristic variable decomposes the first objective function in respect of the optimization parameter into two simple optimization problems, i.e. the function relationship between the intermediate characteristic variable and the optimization parameter, and the second objective function in respect of the intermediate characteristic variable, thus achieving the effects of shortening optimization time and reducing the amount of calculation.

Classes IPC  ?

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

Abrégé

A method for measuring the thermal power of a nuclear reactor core in a DCS, comprising: acquiring reactor core thermal power data and related feature variables of reactor core thermal power; classifying the related feature variables to construct an original feature vector set; performing feature importance sorting on feature subsets by means of an extreme gradient boosting algorithm, and determining an optimal feature subset sequence by means of an extreme learning machine model; acquiring hyperparameters for reactor core thermal power calculation by means of an equilibrium optimizer algorithm; determining an optimal mode component combination of the reactor core thermal power data by means of an integrated empirical mode decomposition algorithm; and calculating the reactor core thermal power by means of a thermal power measurement method based on a thermal balance combination model.

Classes IPC  ?

• G21C 17/00 - SurveillanceTests
• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu
• G06F 17/10 - Opérations mathématiques complexes
• G06N 3/00 - Agencements informatiques fondés sur des modèles biologiques
• G21D 3/00 - Commande des installations à énergie nucléaire

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

A method and apparatus for monitoring a physical field of a nuclear reactor, and a device and a storage medium. The method for monitoring a physical field of a nuclear reactor comprises: on the basis of a response-data splicing rule, splicing real-time response data collected by nuclear reactor detectors of at least two types, so as to obtain a detector measurement value (S201), wherein the nuclear reactor detectors are configured inside and/or outside a nuclear reactor; according to the detector measurement value, determining a reduced basis fitting coefficient (S202); according to a target reduced basis and the reduced basis fitting coefficient, determining a real-time monitoring data set of the nuclear reactor (S203); and according to a physical-field splitting rule corresponding to the response-data splicing rule, determining, from among the real-time monitoring data set, real-time monitoring data of at least two physical fields of the nuclear reactor (S204). By using the method for monitoring a physical field of a nuclear reactor, the physical monitoring efficiency and accuracy of a reactor core of a nuclear reactor can be improved.

Classes IPC  ?

• G21C 17/00 - SurveillanceTests
• G06N 20/00 - Apprentissage automatique
• G06F 17/18 - Opérations mathématiques complexes pour l'évaluation de données statistiques
• G06F 30/20 - Optimisation, vérification ou simulation de l’objet conçu

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

A nuclear reactor core information prediction method and apparatus, a device, and a storage medium. The method comprises: according to, for a nuclear reactor at a present moment, an actual reactor core state, an actual reactor core covariance, and, for at least two particles of the nuclear reactor, an average state error and a state covariance, determining a present particle state and a present particle error of the at least two particles; according to the present particle state and the present particle error of the at least two particles, determining a predicted reactor core state and a predicted reactor core covariance of the nuclear reactor at the next moment; according to the predicted reactor core state and predicted reactor core covariance of the nuclear reactor at the next moment, determining a predicted residual value and a predicted gain value; and according to the predicted residual value, predicted gain value, and the predicted reactor core state and predicted reactor core covariance of the nuclear reactor at the next moment, determining an actual reactor core state and an actual reactor core covariance of the nuclear reactor at the next moment. The invention improves the accuracy of determining nuclear reactor core information.

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"
• G06N 3/0442 - Réseaux récurrents, p. ex. réseaux de Hopfield caractérisés par la présence de mémoire ou de portes, p. ex. mémoire longue à court terme [LSTM] ou unités récurrentes à porte [GRU]

Abrégé

A nuclear power plant system test method and apparatus, and a computer device, a storage medium and a computer program product. The method comprises: according to a parameter transmission requirement of a nuclear power plant system, performing interface configuration on a test terminal (102), so as to obtain a target test terminal corresponding to the parameter transmission requirement of the nuclear power plant system (S202); connecting the nuclear power plant system to the target test terminal (102), so as to obtain a target configuration state (S204); on the basis of the type of data transmitted by the target test terminal (102), determining a test configuration option corresponding to the target configuration state (S206); and on the basis of the test configuration option, executing a test on the nuclear power plant system (S208).

Classes IPC  ?

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

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

An all-solid-state full-static miniature reactor, comprising a reactor core container (1), a graphite matrix (2), which is arranged in the reactor core container (1), and a control protection mechanism (3), which is connected to the graphite matrix (2), wherein the control protection mechanism (3) comprises several shutdown columns (32), the several shutdown columns (32) are arranged in the graphite matrix (2), and when the temperature of the graphite matrix (2) rises to a predetermined temperature, each shutdown column (32) automatically decomposes to obtain a shutdown substance that has negative reactivity feedback to a reactor core reaction, so as to realize passive shutdown of the reactor. The all-solid-state full-static miniature reactor is provided with a plurality of shutdown columns (32), and the shutdown columns (32) decompose at a high temperature relying on the natural characteristics of the shutdown columns (32), thereby realizing natural emergency shutdown, and reducing or canceling the use of active components.

Classes IPC  ?

• G21C 9/02 - Moyens pour effectuer une réduction très rapide du facteur de réactivité dans des conditions défectueuses, p. ex. fusible pour réacteur

Abrégé

A reactor shutdown processing method and apparatus for an offshore nuclear power plant, and a device and a storage medium. The reactor shutdown processing method comprises: acquiring an attitude information set and a motion information set of an offshore nuclear power plant, which sets are collected by at least two identical detection apparatuses, wherein the at least two identical detection apparatuses are configured on the basis of a working condition of an ocean where the offshore nuclear power plant is located, the attitude information set comprises at least two pieces of attitude information, each piece of attitude information comprises a swing angle and/or a tilt angle, the motion information set comprises at least two pieces of motion information, and each piece of motion information comprises an acceleration; and according to a comparison result between a set angle threshold value and attitude information in the attitude information set, and a comparison result between a set acceleration threshold value and motion information in the motion information set, determining whether to perform reactor shutdown processing on the offshore nuclear power plant. By using the present reactor shutdown processing method, the safety impact of a marine working condition on an offshore nuclear power plant can be well determined, thereby improving the safety of the offshore nuclear power plant.

Classes IPC  ?

• G21D 3/04 - Dispositions de sécurité
• G21C 17/00 - SurveillanceTests
• G21D 1/02 - Dispositions d'équipement auxiliaire
• G01D 21/02 - Mesure de plusieurs variables par des moyens non couverts par une seule autre sous-classe

Abrégé

A nuclear reactor, comprising: a reactor vessel (100), a heat exchange device (200), a driving device (300), and a water supply device (400). A liquid coolant and a gas cavity (150) are provided in the reactor vessel (100); the heat exchange device (200) is disposed in the reactor vessel (100); the heat exchange device (200) comprises an outer wall surface (220) and an inner wall surface (210), wherein a first cavity (240) for accommodating a reactor core (500) is formed on the inner wall surface (210), and the outer wall surface (220) and the inner wall surface (210) jointly define a second cavity (250); the driving device (300) has one end connected to the gas cavity (150), and the other end arranged close to the reactor core (500); and the water supply device (400) is communicated with the second cavity (250). According to the nuclear reactor, when an accident condition occurs, the heat at a primary side of the reactor and the reactor core (500) can be discharged, thereby avoiding fuel damage caused by further heating of the reactor core (500).

Classes IPC  ?

• G21C 1/32 - Réacteurs du type intégré, c.-à-d. réacteurs dans lesquels des parties associées de façon fonctionnelle avec le réacteur, mais non essentielles à la réaction, p. ex. des échangeurs de chaleur, sont disposées à l'intérieur de l'enveloppe avec le cœur
• G21C 1/03 - Réacteurs de fission rapides, c.-à-d. réacteurs n'utilisant pas de modérateur refroidis par un réfrigérant non nécessairement pressurisé, p. ex. réacteurs du type piscine
• G21C 15/243 - Cyclage du fluide réfrigérant pour des liquides
• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur
• 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

Abrégé

Provided in the present invention is a bottom pipe base of a nuclear fuel assembly capable of limiting the vibration of fuel rods, the bottom pipe base comprising a base body, a plurality of assembly column and flow channel pipe units and a plurality of main flow channel pipes for liquid to pass through, wherein the base body is provided with a cavity, which is open upwards, and the main flow channel pipes and the assembly column and flow channel pipe units are transversely arranged in the cavity side by side; and the assembly column and flow channel pipe units each comprise a guide pipe assembly column and at least two assembly column side flow channel pipes for the liquid to pass through, the main flow channel pipes and the assembly column side flow channel pipes each comprise limiting inclined surface arranged at the side surface of the upper end, and the limiting inclined surfaces enclose a recessed structure, which is open upwards and adapted for receiving the ends of external fuel rods. The bottom pipe base of a nuclear fuel assembly of the present invention can achieve the limiting effect on the ends of fuel rods, and further reduces flow-induced vibration of the fuel rods; and a longitudinal through structure of the main flow channel pipes and the assembly column side flow channel pipes can be adapted for a cooling liquid to pass through, thereby making the axial flow of the cooling liquid more uniform and reducing the cross flow between the fuel rods, such that the flow-induced vibration of the fuel rods is reduced.

Classes IPC  ?

• G21C 3/33 - Moyens pour supporter ou suspendre des éléments dans le faisceauMoyens faisant partie du faisceau pour l'insérer dans le cœur ou l'en extraireMoyens de couplage de faisceaux adjacents
• G21C 3/322 - Moyens pour influencer l'écoulement du réfrigérant à travers ou autour des faisceaux

Abrégé

The present invention provides a nuclear fuel assembly bottom nozzle comprising a constraint anti-foreign matter plate, comprising a base body and an anti-foreign matter plate. The anti-foreign matter plate is horizontally arranged on the upper side of the base body and comprises multiple partition plates; the multiple partition plates intersect to define multiple through holes; the through holes longitudinally pass through the anti-foreign matter plate; the anti-foreign matter plate is further provided with multiple constraint holes that match the bottom ends of external fuel rods and allow the bottom ends of the external fuel rods to insert; the inner walls of the constraint holes are inclined, and the diameter of each constraint hole gradually decreases from top to bottom. The constraint holes have an effect of constraining the vibration of the fuel rods, so as to reduce flow-induced vibration of the fuel rods or the vibration of a fuel assembly caused by an external force, especially the vibration of lower sections of the fuel rods, and achieves a protective effect of the bottom nozzle on the ends of the fuel rods, thereby avoiding abrasion between the ends of the fuel rods and the bottom nozzle caused by the flow-induced vibration. The anti-foreign matter plate is arranged on the upper side of the base body, so that the anti-foreign matter plate can match base bodies of various structures.

Classes IPC  ?

• G21C 3/33 - Moyens pour supporter ou suspendre des éléments dans le faisceauMoyens faisant partie du faisceau pour l'insérer dans le cœur ou l'en extraireMoyens de couplage de faisceaux adjacents

Abrégé

Disclosed in the present invention are a fuel rod clamping ring and a fuel assembly supporting grid. The fuel rod clamping ring is arranged in a supporting grid to clamp a fuel rod inserted into the supporting grid. The fuel rod clamping ring comprises a clamping ring body, a plurality of clamping portions arranged on the clamping ring body in a protruding mode so as to clamp the fuel rod penetrating through the clamping ring body, and a plurality of connecting portions arranged on the clamping ring body and connected to the supporting grid. According to the present invention, the clamping portions of the fuel rod clamping ring protrude out of the clamping ring body, facilitating formation of stable surface contact between the clamping portions and the fuel rod, so that local abrasion of the surface of the fuel rod by the clamping portions can be reduced; the clamping ring can be integrally placed in the supporting grid only by connecting the connecting portions to the supporting grid, so that the clamping ring is easy to machine and assemble .

Classes IPC  ?

• G21C 3/356 - Grilles d'espacement munies d'organes de support des éléments combustibles
• G21C 3/34 - Grilles d'espacement
• G21C 3/33 - Moyens pour supporter ou suspendre des éléments dans le faisceauMoyens faisant partie du faisceau pour l'insérer dans le cœur ou l'en extraireMoyens de couplage de faisceaux adjacents

Abrégé

A method and apparatus for determining a flow in a primary loop of a reactor, a computer device, a storage medium, and a computer program product. The method comprises: acquiring the current rotating speed of a main pump (104) in a primary loop of a reactor (S201); according to the current rotating speed, a rated rotating speed of the main pump (104), a flow to be solved in the primary loop of the reactor, and a first lift mapping function, determining a lift at the rated rotating speed (S202); according to the current rotating speed, the lift at the rated rotating speed, and a first relationship, determining a first lift at the current rotating speed (S203); according to said flow and a second lift mapping function, determining a second lift at the current rotating speed (S204); and determining the value of said flow according to the first lift and the second lift (S205). According to the method, the value of a flow to be solved is calculated on the basis of the flow of a main pump (104) at a rated rotating speed, thereby avoiding errors caused by flowmeter measurement and differential pressure measurement, and achieving high accuracy of determination results.

Classes IPC  ?

• G21C 17/032 - Mesure ou surveillance du débit de réfrigérant

Abrégé

The present application relates to an exciter rotor dismounting/mounting device and method, for use in disassembling or assembling an exciter rotor (810) and a generator rotor (820) which cooperate with each other. The device comprises a weight measuring assembly (200). The weight measuring assembly (200) comprises a first weight measuring assembly (210) and a second weight measuring assembly (220) arranged at an interval in the axial direction of the exciter rotor (810). The first weight measuring assembly (210) and the second weight measuring assembly (220) respectively abut against different specific positions of the exciter rotor (810), so as to cooperate with each other to support the different specific positions of the exciter rotor (810) in the axial direction, and respectively measure the values of pressures, borne by the first weight measuring assembly (210) and the second weight measuring assembly (220), from the self-exciter rotor (810). The first weight measuring assembly (210) and the second weight measuring assembly (220) can move in the vertical direction so as to adjust the levelness of the exciter rotor (810).

Classes IPC  ?

• H02K 15/02 - Procédés ou appareils spécialement adaptés à la fabrication, l'assemblage, l'entretien ou la réparation des machines dynamo-électriques des corps statoriques ou rotoriques

Abrégé

The present application relates to the technical field of nuclear reactors, and in particular, to a reactor control method and apparatus, a computer device and a storage medium. The method comprises: determining a target rod group from candidate rod groups according to a steam turbine load and nuclear power of a reactor; determining action speeds and directions of rod groups according to the current temperature of a reactor coolant and a coolant temperature setting value corresponding to the steam turbine load; and controlling an action of the target rod group according to the action speeds and directions of the rod groups. The present application simplifies the adjustment process of rod groups, and optimizes the power adjustment process.

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 7/36 - Circuits de commande
• G21C 7/06 - 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

Abrégé

An aerosol sedimentation mechanism research device, comprising: a test tube (100), an aerosol generator (200), an air compressor (300), a cooling assembly (400) and an exhaust treatment box (500). The test tube (100) is provided with an inlet end and an outlet end which are communicated with each other, and aerosol concentration testers are provided at both the inlet end and the outlet end; the aerosol generator (200) is used for preparing an aerosol, and the aerosol generator (200) can be selectively in communication with the inlet end; the air compressor (300) can be selectively in communication with the inlet end; the cooling assembly (400) comprises a cooler (410) and a water supply tank (420) which are communicated with each other, the water supply tank (420) being filled with a cooling liquid, the cooler (410) being mounted on the outer tube wall of the test tube (100), and the cooling liquid in the water supply tank (420) being capable of flowing between the cooler (410) and the water supply tank (420) to regulate the wall temperature of the test tube (100); the inlet end is in communication with the exhaust treatment box (500); the exhaust treatment box (500) is used for accommodating a treatment solution; the aerosol in the test tube (100) can be dissolved in the treatment solution. Further disclosed is an operation method for the aerosol sedimentation mechanism research device.

Classes IPC  ?

• G01N 15/06 - Recherche de la concentration des suspensions de particules
• G01N 15/00 - Recherche de caractéristiques de particulesRecherche de la perméabilité, du volume des pores ou de l'aire superficielle effective de matériaux poreux

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

An apparatus and method for measuring a melting point of an ultrahigh-temperature material. The apparatus for measuring the melting point of the ultrahigh-temperature material comprises: a closed measurement chamber (10), a sample stage (20) provided in the measurement chamber (10) and used for placing a sample (100) to be measured, a plasma system connected to at least one side of the measurement chamber (10) and used for generating a plasma jet as a heat source to heat said sample (100), and a non-contact temperature measurement unit used for measuring the temperature of said sample (100); wherein the measurement chamber (10) is provided with at least one temperature measurement window (11) corresponding to the non-contact temperature measurement unit, and at least one observation window (12) for photographing and recording. According to the apparatus, said sample (100) whose melting point is to be measured is heated by means of plasma heating, so that said sample (100) can be melted by means of surface heating, and the melting point of said sample (100) is obtained by means of temperature measurement, and the problems of small spot size and difficulty in extracting temperature information when measuring the melting point with a laser are solved. By synchronously measuring the deformation process and temperature of said sample (100)when heated by plasma, the measurement precision is improved.

Classes IPC  ?

• G01N 25/4 -

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 governor and a governor control method. The governor comprises: two control circuits, each control circuit comprising two processors, the four processors being connected to each other, and the types of two processors in a same control circuit being different. When a processor in a master control circuit (11) fails, the failed processor in the master control circuit (11) sends synchronization information to a first target processor in a slave control circuit (12), so that the first target processor is connected to a second target processor to form a target data transmission channel so as to achieve data transmission. The first target processor is a processor in the slave control circuit (12) having the same type as the failed processor, and the types of the second target processor and the first target processor are different. By using the governor, a system failure caused by a single point of failure can be avoided.

Classes IPC  ?

• G05B 9/03 - Dispositions de sécurité électriques avec une boucle à canal multiple, c.-à-d. systèmes de commande redondants
• G05D 13/62 - Commande de la vitesse linéaireCommande de la vitesse angulaireCommande de l'accélération ou de la décélération, p. ex. d'une machine motrice caractérisée par l'utilisation de moyens électriques, p. ex. l'emploi de dynamos-tachymétriques, l'emploi de transducteurs convertissant des valeurs électriques en un déplacement

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

The present application relates to a multi-fidelity network construction method and apparatus for a nuclear reactor simulation test. The method comprises: acquiring a first fidelity network according to first fidelity data of a sample nuclear reactor, and acquiring at least one second fidelity network according to second fidelity data of the sample nuclear reactor (102); training the at least one second fidelity network by using the second fidelity data to obtain at least one trained second fidelity network (104); and combining the at least one trained second fidelity network with the first fidelity network to obtain a multi-fidelity network, and training the multi-fidelity network by using the first fidelity data to obtain a trained multi-fidelity network, wherein the trained multi-fidelity network is used for performing a simulation test on a target nuclear reactor (106). By using the method, a final simulation result can be output according to the coupling between different fidelity data, thereby improving the simulation efficiency while ensuring the simulation precision.

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 18/214 - Génération de motifs d'entraînementProcédés de Bootstrapping, p. ex. ”bagging” ou ”boosting”
• G06N 3/04 - Architecture, p. ex. topologie d'interconnexion
• G06N 3/08 - Méthodes d'apprentissage

Abrégé

Disclosed in the present invention are a fuel assembly fixing device and a fuel assembly fixing method. The fuel assembly fixing device is used for detachably connecting a fuel assembly to a lower core plate, and comprises a limiting member, an elastic member, elastic pieces, and a pipe leg connected to the fuel assembly. The limiting member is arranged inside the pipe leg, and the limiting member comprises a limiting section pipe wall and an unlocking section pipe wall having an outer diameter less than that of the limiting section pipe wall. The limiting section pipe wall is closer to the inner bottom of the pipe leg than the unlocking section pipe wall. The elastic member has one end in contact with the limiting section pipe wall, and the other end in contact with the inner bottom of the pipe leg. A first end of each elastic piece is connected to the inner side wall of the pipe leg. The fuel assembly fixing device has an unlocking state and a locking state. When the fuel assembly fixing device is switched from the unlocking state to the locking state, the limiting section pipe wall returns to be in contact with the elastic piece and presses the free end of the elastic piece to move towards the outer side wall of the pipe leg, and a second end of the elastic piece extends out of the pipe leg and is locked to the bottom of the lower core plate. The fuel assembly fixing device of the present invention can maintain the positioning of the fuel assembly on the lower core plate, thereby preventing the fuel assembly from floating upwards.

Classes IPC  ?

• G21C 3/332 - Supports pour grilles d'espacement

Abrégé

The present application relates to a 5G-based abnormity diagnosis method and apparatus for a nuclear power device, and a computer device. The method comprises: obtaining sensing data of a nuclear power device at each moment in a preset time period, performing dimension reduction processing on the sensing data, and calculating a maximum value and a standard deviation of the sensing data after dimension reduction processing; and if any one of the following conditions is met, determining that the sensing data after dimension reduction processing is abnormal data, and obtaining, according to the abnormal data, a diagnosis result that the nuclear power device is abnormal, wherein said following conditions comprise that the maximum value is greater than a preset maximum value threshold, and the standard deviation is greater than a preset standard deviation threshold.

Classes IPC  ?

• G06F 11/34 - Enregistrement ou évaluation statistique de l'activité du calculateur, p. ex. des interruptions ou des opérations d'entrée–sortie

Abrégé

A split-type nuclear fuel pellet structure, and a fuel rod having an SiC composite cladding. The split-type nuclear fuel pellet structure comprises a pellet (1) and an elastic member (2), wherein a cavity (11) for accommodating the elastic member (2) is provided in the middle of the pellet (1); and the pellet (1) comprises at least two pellet sections (10) radially fitting each other, and the elastic member (2) is accommodated in the cavity (11) and abuts against the pellet sections (10). In the split-type nuclear fuel pellet structure, the pellet (1) is in a tight fit with a nuclear fuel cladding tube (3) without gaps, which can ensure the uniform circumferential heat transfer of the fuel rod. The pellet (1) having the cavity (11) in the middle can further reduce the operating peak temperature of the pellet (1), and the cavity (11) can accommodate more fission gas, such that the internal pressure of the fuel rod can be reduced.

Classes IPC  ?

• G21C 3/58 - Combustible pour réacteur solide
• G21C 3/06 - EnveloppesChemises

Abrégé

A vision-guided intelligent fastening bolt inspection device, comprising: a frame (11); a main buoyancy body (1), which is mounted on the frame (11) and serves as a main buoyancy source of the inspection device; propellers (3), used for controlling the movement direction of the inspection device; an ultrasonic inspection module (5); and an underwater self-stabilization moving platform (4), which comprises an upper nut sliding block (403) and a lower nut sliding block (411) which are opposite in movement direction, a balancing weight (408), and a self-stabilization driving part used for driving the upper nut sliding block (403) and the lower nut sliding block (411) to move in opposite directions, wherein the ultrasonic inspection module (5) is fixedly connected to the upper nut sliding block (403), and the balancing weight (408) is fixedly connected to the lower nut sliding block (411), or the ultrasonic inspection module (5) is fixedly connected to the lower nut sliding block (411), and the balancing weight (408) is fixedly connected to the upper nut sliding block (403).

Classes IPC  ?

• B63C 11/52 - Outillage spécialement adapté au travail sous l'eau, non prévu ailleurs
• B63G 8/16 - Commande de l'assiette ou de la profondeur par emploi direct des hélices ou d'un système à réaction
• B63G 8/26 - Dispositifs de réglage de l'assiette
• B63G 8/39 - Disposition des instruments de veille sonique, p. ex. basse fréquence, sonar

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

A transient analysis method for a nuclear power plant accident causing pellet-cladding interaction. The method comprises: acquiring the same xenon strategy for a nuclear thermal system program and a reactor core computing program (S100); acquiring an initial-moment axial power deviation of the nuclear thermal system program and an initial-moment axial power deviation of the reactor core computing program (S200); when the initial-moment axial power deviation of the reactor core computing program is consistent with the initial-moment axial power deviation of the nuclear thermal system program, acquiring a neutronics parameter of the reactor core computing program and a neutronics parameter of the nuclear thermal system program (S300); correcting the neutronics parameter of the nuclear thermal system program to be consistent with the neutronics parameter of the reactor core computing program (S400); according to the corrected neutronics parameter of the nuclear thermal system program, acquiring a boundary condition of a transient accident process that causes pellet-cladding interaction (S500); and obtaining the power distribution of the reactor core computing program in the transient accident process (S600).

Classes IPC  ?

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

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 application relates to a power tilt prediction method and apparatus, a device, a storage medium, and a program product. The method comprises: first obtaining the measurement activity under any burnup in a reactor core operation process; then carrying out iterative solution according to the measurement activity and a pre-established objective function to obtain an optimal water gap parameter; and finally, carrying out power tilt prediction on the basis of the optimal water gap parameter to obtain a target tilt factor. By using the method, the change trend of the quadrant power tilt along with burnup and a power level can be obtained, and then a quadrant power tilt factor of a reactor core can be quantitatively predicted; the prediction effect is good, the arbitrariness of manual adjustments is avoided, the reliability is high, and the operation plan of a unit can be intervened in advance, thereby improving the operation safety of the unit.

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

The present application relates to a nuclear reactor fault determination method, an apparatus, a device, a storage medium, and a product. The method comprises: first, acquiring operating state data when a target event in a nuclear reactor has a fault; then by means of a fault inference model, analyzing the operating state data to acquire fault probabilities of a plurality of candidate events associated with target event logic; and, according to the fault probability of each candidate event, determining a fault determination result of the nuclear reactor. The fault inference model is constructed on the basis of operating state data of historical fault events of the nuclear reactor and conditional fault probabilities between the historical fault events. The present method can be used for accurately acquiring fault determination results of nuclear reactors.

Classes IPC  ?

• G06Q 10/20 - Administration de la réparation ou de la maintenance des produits
• G06Q 50/06 - Fourniture d’énergie ou d’eau
• G06N 5/045 - Explication d’inférenceIntelligence artificielle explicable [XAI]Intelligence artificielle interprétable
• G06N 7/01 - Modèles graphiques probabilistes, p. ex. réseaux probabilistes
• G21D 3/06 - Dispositions de sécurité réagissant à des défaillances à l'intérieur de l'installation

Abrégé

A support grid for critical heat flux (CHF) of a nuclear fuel assembly, and a method for adjusting the bending dimension of a fuel rod (4). The support grid for CHF of a nuclear fuel assembly comprises a support grid body (1), a supporting structure (2) for supporting the fuel rod (4), and a support member (5) that is in contact with the fuel rod (4) and provides a bending fulcrum. The support grid body (1) is provided with a mounting structure (11); a part of the mounting structure (11) extends out of the end surface of the support grid body (1); the supporting structure (2) is connected to the mounting structure (11), and a part of the supporting structure (2) extends out of the end surface of the support grid body (1); the support member (5) is arranged at one end of the supporting structure (2) and located outside the end surface of the support grid body (1). According to the support grid for CHF of a nuclear fuel assembly, a bending portion of the fuel rod (4) can be located outside the end surface of the support grid body (1), so that the influences of the presence of the support grid on the area of a flow channel and the flow velocity of a fluid are avoided, thereby making the obtained experimental result of a CHF test of the nuclear fuel assembly more reliable.

Classes IPC  ?

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

Abrégé

The present application relates to a matrix, a control rod, a reactor control assembly, and a system. The matrix (120) of the present application is adapted to an absorption core (110) and a cladding (130) of the control rod (10). The matrix (110) is provided with a core recess (121), and the core recess (121) is used for accommodating the absorption core (110). In use, the absorption core (110) is located in the core recess (121), and the matrix (120) is located in the cladding (130).

Classes IPC  ?

• B23K 31/02 - Procédés relevant de la présente sous-classe, spécialement adaptés à des objets ou des buts particuliers, mais non couverts par un seul des groupes principaux relatifs au brasage ou au soudage
• G21C 7/10 - Structures des éléments de commande

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

An online test apparatus and method for an oxygen sensor in a lead-bismuth reactor. The online test apparatus for an oxygen sensor in a lead-bismuth reactor comprises: a lead-bismuth container (10), which contains liquid lead and bismuth and is used for an oxygen sensor to be tested (40) to be inserted therein, and an electrochemical workstation (20) and a data processing system (30), wherein a reference electrode and a counter electrode of the electrochemical workstation (20) are both connected to a reference electrode of said oxygen sensor (40), and a sensing electrode and a working electrode of the electrochemical workstation (20) are connected to the liquid lead and bismuth via leads (23, 24), respectively. By means of the response of said oxygen sensor (40) to an alternating-current voltage applied within a specific frequency range, the online test apparatus for an oxygen sensor in a lead-bismuth reactor realizes online monitoring of the operation state of said oxygen sensor (40) by means of electrochemical impedance spectroscopy.

Classes IPC  ?

• G01N 27/411 - Cellules et sondes avec des électrolytes solides pour la recherche ou l'analyse de métaux liquides
• G21C 17/02 - Dispositifs ou dispositions pour la surveillance du réfrigérant ou du modérateur
• G21C 19/07 - Râteliers de stockagePiscines de stockage

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

A highly thermally conductive nuclear fuel element, comprising a closed silicon carbide cladding (10), and a fuel pellet (20) and a thermally conductive network (30) that are provided in the silicon carbide cladding (10). A gap cavity (40) for being filled with an inert gas is reserved between the silicon carbide cladding (10) and the fuel pellet (20); and the thermally conductive network (30) is located in the gap cavity (40), and is in thermally conductive connection with the outer surface of the fuel pellet (20) and the inner surface of the silicon carbide cladding (10), respectively. According to the highly thermally conductive nuclear fuel element, the thermally conductive network (30) is arranged between the silicon carbide cladding (10) and the fuel pellet (20), thereby improving the heat transfer characteristics of the nuclear fuel element under normal operation conditions of a reactor, and avoiding the generation of contact by-products.

Classes IPC  ?

• G21C 3/08 - EnveloppesChemises munis de moyens externes pour favoriser l'échange de chaleur, p. ex. ailettes, déflecteurs, cannelures
• G21C 3/22 - Éléments combustibles avec un matériau fissile ou surrégénérateur en contact avec le réfrigérant
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages

Abrégé

Disclosed in the present invention are a high-density silicon carbide composite material cladding and a preparation method therefor. The preparation method comprises the following steps: S1, preparing a slurry; S2, processing and forming the slurry into a silicon carbide cladding green body; and S3, degreasing and sintering the silicon carbide cladding green body to form a high-density silicon carbide cladding. In the preparation method for the high-density silicon carbide composite material cladding in the present invention, a silicon carbide fiber powder mixture is coordinated with an organic solvent, etc., to prepare a slurry, the slurry is then processed and formed into a cladding green body, and degreasing and sintering treatments are performed to prepare a nanoscale silicon carbide cladding with a super length-diameter ratio, such that the density of the silicon carbide cladding is improved, pore defects are reduced, and harmful byproducts of the silicon carbide cladding are decreased; a preparation process for the cladding is simplified, and the production efficiency is improved; and the preparation cost is reduced.

Classes IPC  ?

• C04B 35/80 - Fibres, filaments, "whiskers", paillettes ou analogues
• C04B 35/565 - Produits céramiques mis en forme, caractérisés par leur compositionCompositions céramiquesTraitement de poudres de composés inorganiques préalablement à la fabrication de produits céramiques à base de non oxydes à base de carbures à base de carbure de silicium
• C04B 35/622 - Procédés de mise en formeTraitement de poudres de composés inorganiques préalablement à la fabrication de produits céramiques

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 primary loop water chemical determination method, comprising: acquiring the concentration of unionized boric acid and of unionized lithium hydroxide in primary loop water at the current measurement temperature, and the saturation solubility in the primary loop of each of the remaining metal elements water except for lithium element in the primary loop water (S11); constructing a first function according to ion charge conservation, an ion product formula of boric acid and lithium hydroxide in the primary loop water and the concentration of unionized boric acid and of unionized lithium hydroxide, wherein the first function is used for characterizing the corresponding relationship between the total concentration of lithium hydroxide and the pH value (S12); and calculating the current pH value of the primary loop water according to the first function and the current measured total concentration of lithium hydroxide (S13).

Classes IPC  ?

• G01N 33/18 - Eau
• G06F 17/15 - Calcul de fonction de corrélation

Abrégé

The present application relates to a cladding tube blast simulation method and apparatus, a device, a storage medium, and a product. The method comprises: acquiring pressurized blast experiment parameters of a cladding tube; according to the pressurized blast experiment parameters, applying pressure loads to a virtual model which corresponds to the cladding tube, and carrying out cladding tube blast simulation experiments, so as to generate pressurized blast simulation parameters; and analyzing the pressurized blast simulation parameters, so as to generate a pressurized blast simulation result. In the present solution, a real pressurized blast experiment is replaced with a virtual experiment method, various operating conditions of the real pressurized blast experiment are simulated by means of various corresponding pressurized blast experiment parameters, thereby avoiding the waste of cladding tube materials in a large number of experiments carried out in the real pressurized blast experiment. In this way, the cost of pressurized blast experiments is reduced, and the economy of the pressurized blast experiments is improved.

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

The present application relates to the technical field of nuclear fuels. Embodiments of the present application provide a silicon carbide composite connecting device (100) configured to connect a cladding tube (210) and two end plugs (220). Mounting holes (211) are respectively provided in two opposite ends of the cladding tube (210) in the axial direction of the cladding tube (210), and each mounting hole (211) is configured to mount an end plug (220). The silicon carbide composite connecting device (100) comprises: a device body (110), wherein the device body (110) is provided with a first accommodating cavity (111), the first accommodating cavity (111) is configured to accommodate gas having a preset pressure, and the cladding tube (210) and one end plug (220) connected to the cladding tube (210) can extend into the first accommodating cavity (111); and a heating structure, wherein the heating structure is provided on the device body (110) and is configured to heat positions at which the cladding tube (210) and the end plugs (220) are connected.

Classes IPC  ?

• G21C 21/02 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages
• G21C 3/10 - Obturateurs d'extrémités
• B23K 9/16 - Soudage ou découpage à l'arc utilisant des gaz de protection
• B23K 1/00 - Brasage ou débrasage
• B23K 26/12 - Travail par rayon laser, p. ex. soudage, découpage ou perçage sous atmosphère particulière, p. ex. dans une enceinte
• B23K 15/10 - Soudage ou découpage par faisceau d'électrons n'ayant pas lieu sous vide
• C04B 37/00 - Liaison des articles céramiques cuits avec d'autres articles céramiques cuits ou d'autres articles, par chauffage

Abrégé

The present application particularly relates to a defect identification method and apparatus, a computer device, and a storage medium. The method comprises: determining, by means of an image reconstruction network in a defect identification model, whether an image to be identified comprises a defect region, and a discrete distribution value corresponding to said image; and if said image contains the defect region, determining, by means of a classification network in the defect identification model, a target defect category of said image according to the discrete distribution value corresponding to said image.

Classes IPC  ?

• G06T 7/00 - Analyse d'image

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

A nuclear power safety injection tank (10), comprising a tank body (11), an isolation structure (12) and a regulating cylinder (13), wherein the tank body (11) is provided with a pressurizing port (111) at an upper end, and is provided with a flow-through port (112) at a lower end; the isolation structure (12) is arranged on an inner wall surface of the tank body (11) such that the space in the tank body (11) close to an inner side face is partitioned in a height direction so as to block liquid in the tank body (11); and the regulating cylinder (13) is vertically arranged at the bottom of the tank body (11), and has a lower end connected to the bottom of the tank body (11), regulating holes (131) are distributed in the regulating cylinder (13) in the height direction, and the flow-through port (112) communicates with the interior of the regulating cylinder (13).

Classes IPC  ?

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

Abrégé

An adjustment apparatus, for use in adjusting a rotating speed sensor (4) of a main pump spindle (2). A probe (1) is arranged on the outer wall surface of the main pump spindle (2). A sensing groove (41) penetrating in the circumferential direction is formed in the end of the rotating speed sensor (4) opposite to the main pump spindle (2). The sensing groove (41) is used for the probe (1) to rotate through. The adjustment apparatus comprises a mounting bracket (5) arranged on a main pump housing and used for mounting the rotating speed sensor (4); and a probe simulator (32) comprising a positioning head (B) for positioning the probe (1) and a measurement assembly (324), wherein the positioning head (B) rotates to pass through the sensing groove (41), and the measurement assembly (324) is used for acquiring a distance between the positioning head (B) and the axial inner side wall of the sensing groove (41) and a distance between the positioning head (B) and the radial bottom wall of the sensing groove (41). The position of the rotating speed sensor (4) on the mounting bracket (5) is adjustable, the position of the rotating speed sensor relative to the probe (1) and a positioning member (321) can be adjusted, the accurate relative position between the rotating speed sensor (4) and the probe (1) is guaranteed, and normal rotating speed detection of the rotating speed sensor (4) on the probe (1) is achieved. Also provided is an adjustment method.

Classes IPC  ?

• G01P 21/02 - Essai ou étalonnage d'appareils ou de dispositifs couverts par les autres groupes de la présente sous-classe de tachymètres

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

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

A ceramic joining device, comprising a working furnace (100), a clamping mechanism (130), a vacuum interface (140), and a heater (150). A furnace cavity (111) is formed in the working furnace (100) and is used for accommodating a target material (300). The clamping mechanism (130) is located in the furnace cavity (111) and is used for fixing samples to be joined (400). The vacuum interface (140) is communicated with the furnace cavity (111) and is configured to be joined to a vacuum generator, so that the furnace cavity (111) is in a vacuum environment. The heater (150) is joined to the working furnace (100) and is used for heating the target material (300) in the furnace cavity (111) to a preset temperature. When the vacuum degree in the furnace cavity is lower than the saturation vapor pressure of the target material, the target material at the preset temperature can be gathered to said samples in the form of steam, so that a gas permeation reaction occurs in a gap of a ceramic to form a joining layer, and the joining between said two samples is realized, without filling a joining position with a joining material in advance. Therefore, the processing steps of welding of said samples are simplified, and the processing difficulty is reduced.

Classes IPC  ?

• B23K 37/00 - Dispositifs ou procédés auxiliaires non spécialement adaptés à un procédé couvert par un seul des autres groupes principaux de la présente sous-classe
• B23K 1/008 - Brasage dans un four

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

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 device for testing steam turbine valve characteristics, comprising a case, wherein a control card (1), an acquisition card (2) connected to the control card (1), an acquisition interface (3) connected to the acquisition card (2) and a pressure sensor in a hydraulic servomotor, and an industrial personal computer (4) connected to the control card (1) and the acquisition card (2) are provided in the case. According to the device, a worker can comprehensively and intuitively monitor the operation of a proportional valve and perform fault diagnosis, so that a fault can be predicted in advance, and a fault handling person is prompted to take a corresponding measure, and during maintenance of a control system for a steam turbine valve, the control system can be replaced with the device to control the proportional valve, so that the reliability of a steam turbine valve system is improved.

Classes IPC  ?

• G01R 31/00 - Dispositions pour tester les propriétés électriquesDispositions pour la localisation des pannes électriquesDispositions pour tests électriques caractérisées par ce qui est testé, non prévues ailleurs
• F16K 37/00 - Moyens particuliers portés par ou sur les soupapes ou autres dispositifs d'obturation pour repérer ou enregistrer leur fonctionnement ou pour permettre de donner l'alarme

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

Disclosed in the present invention are a mutual-inductance type liquid metal leakage monitoring apparatus and the use thereof. The mutual-induction type liquid metal leakage monitoring apparatus comprises a detection sensor used for being inserted into a leakage monitoring position. The detection sensor comprises a bendable tube body, a probe assembly arranged at one end portion of the tube body, and a cable passing through the internal portion of the tube body and connected to the probe assembly. The probe assembly comprises a protective sleeve connected to the tube body, a coil skeleton provided in the protective sleeve, and a primary coil and a secondary coil which are successively wound on the coil skeleton, joints of the primary coil and the secondary coil respectively penetrating through the coil skeleton and being connected to the cable. The mutual-inductance type liquid metal leakage monitoring apparatus of the present invention performs monitoring on the basis of the electromagnetic induction principle, and does not need to make contact with mediums under monitoring, thereby avoiding false triggering problems caused by impurity pollution, insulation aging, etc., and effectively reducing the false alarm rate. The mutual-inductance type liquid metal leakage monitoring apparatus of the present invention has the advantages of high reliability, high sensitivity, etc.

Classes IPC  ?

• G01M 3/40 - Examen de l'étanchéité des structures ou ouvrages vis-à-vis d'un fluide par utilisation de moyens électriques, p. ex. par observation de décharges électriques
• G01V 3/11 - Prospection ou détection électrique ou magnétiqueMesure des caractéristiques du champ magnétique de la terre, p. ex. de la déclinaison ou de la déviation fonctionnant au moyen de champs magnétiques ou électriques produits ou modifiés par les objets ou les structures géologiques, ou par les dispositifs de détection en utilisant des cadres inducteurs pour la détection d'objets conducteurs, p. ex. d'armes à feu, de câbles ou de tuyaux
• G01N 27/90 - 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 en utilisant les courants de Foucault

Abrégé

The present invention relates to a generator deformation monitoring method. The method comprises: S1. prior to a generator being pressurized, determining a monitoring position and a monitoring station; S2. by means of a monitoring apparatus arranged at the monitoring station, obtaining position information of the monitoring position, and using same as first monitoring position information; S3. constructing a standard spatial coordinate system, and converting the first monitoring position information to first coordinate data under the standard spatial coordinate system; S4. following generator pressurization, again obtaining position information of the monitoring position, and using same as second monitoring position information; S5. fitting the second monitoring position information into the standard spatial coordinate system, and obtaining second coordinate data; S6. comparing the first coordinate data against the second coordinate data to obtain a deformation monitoring result. The present invention effectively obtains changes to a generator monitoring position in a three-dimensional space, grasps generator deformation patterns before and after pressurization, and collects numerous types of measurement data with relatively little on-site intervention, ensuring the normal operation of the unit, and improving device reliability.

Classes IPC  ?

• G01B 11/00 - Dispositions pour la mesure caractérisées par l'utilisation de techniques optiques

Abrégé

Disclosed in the present invention are a generator iron core fault detection method and system, for determining whether a fault is present at a suspected fault point of a generator iron core. The method comprises: performing local magnetic circuit setting on a suspected fault point, and building a test magnetic circuit; performing excitation processing on the test magnetic circuit, to make the test magnetic circuit enter a predetermined saturated state for a preset duration; and collecting thermal imaging information of the test magnetic circuit, and generating a fault determination result according to the thermal imaging information. Compared with a conventional iron loss method, implementing the present invention can greatly reduce the power required by a test power supply, which can be easily met in a nuclear power plant, without the need to transport a large quantity of heavy test devices. Moreover, the method is simple and efficient, whether an iron core has a fault point can be accurately determined, and the iron loss test duration, workload and cost are effectively reduced.

Classes IPC  ?

• G01R 31/34 - Tests de machines dynamoélectriques
• G01R 19/00 - Dispositions pour procéder aux mesures de courant ou de tension ou pour en indiquer l'existence ou le signe
• G01J 5/48 - ThermographieTechniques utilisant des moyens entièrement visuels

Abrégé

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

Classes IPC  ?

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

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 front water inlet pre-intercepting and filtering device for a pump station of a nuclear power plant, comprising: a frame (10), annular chains (20) being respectively provided on two sides of the frame (10), and the bottom ends of the annular chains (20) being spaced apart from the bottom of a water flow channel; mesh plates (30) continuously and fully distributed on the two annular chains (20), and the two sides of each mesh plate (30) being fixedly connected the two annular chains (20), respectively; and a driving device comprising a motor (110), a main shaft (116) mounted on the frame (10) by means of bearing seats (114), and two chainwheels (118) provided on the main shaft (116), the motor (110) driving the main shaft (116) to rotate, and driving, by means of the engagement of the two chainwheels (118) and the two annular chains (20), the two annular chains (20) and the mesh plates (30) fixedly connected to the two annular chains (20) to perform reciprocating motion. In the front water inlet pre-intercepting and filtering device for a pump station of a nuclear power plant, the bottom ends of the annular chains (20) and the bottom ends of the mesh plates (30) are all spaced apart from the bottom of the water flow channel, thereby guaranteeing a safe water passing channel under an extreme working condition that all meshes (302) are blocked, performing non-full-section water passing pre-interception, increasing the flow passing section, effectively pre-intercepting and filtering marine organisms, and guaranteeing safe and stable operation of the unit.

Classes IPC  ?

• B01D 33/333 - Filtres avec éléments filtrants mobiles au cours de l'opération de filtration avec des éléments filtrants individuels se déplaçant suivant un circuit fermé
• B01D 33/80 - Accessoires
• B01D 33/48 - Rinçage par chasse, p. ex. rafale ascendante d'air à contre-courant
• G06F 17/12 - Opérations mathématiques complexes pour la résolution d'équations d'équations simultanées

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 suction type sampling system for a molten metal medium in a reactor vessel, comprising a suction apparatus (2), a sampling pipeline (3) and a complete sampling box (4). The sampling pipeline (3) is provided with a first port (31) and a second port (32) opposite to and communicated with each other; the first port (31) is connected to the suction apparatus (2); the second port (32) is connected to the reactor vessel; a sampling valve is provided on the sampling pipeline (3); and the sampling valve is located between the first port (31) and the second port (32) and is configured to be capable of communicating or disconnecting the first port (31) and the second port (32). The complete sampling box (4) is provided with a sealing cavity, and the sampling valve is sealed in the sealing cavity. The suction apparatus (2) is used for suctioning away the molten metal medium in the reactor vessel by means of the sampling pipeline (3), or enabling the residual molten metal medium in the sampling pipeline (3) to flow back into the reactor vessel.

Classes IPC  ?

• G01N 1/14 - Dispositifs d'aspiration, p. ex. pompesDispositifs d'éjection
• G21C 17/00 - SurveillanceTests

Abrégé

A flexible robot (1) with a flexible arm (2). The flexible robot comprises the flexible arm (2) and a turning structure (4) configured to drive the flexible arm (2) to turn, wherein the flexible arm (2) comprises an inner retraction portion (21) located on an inner side, and an outer extending portion (22) located on an outer side of the inner retraction portion (21); a cavity (12) is formed between the inner retraction portion (21) and the outer extending portion (22); after a medium is filled into the cavity (12), an end portion of the inner retraction portion (21) extends outwards to form the outer extending portion (22); and the turning mechanism (4) is provided at the end portion, which changes to the outer extending portion (22), of the inner retraction portion (21). With regard to the flexible robot (1), after the medium is filled into the cavity between the inner retraction portion (21) and the outer extending portion (22), the end portion of the inner retraction portion (21) extends outwards to form the outer extending portion (22), so that the flexible arm (2) continuously extends forwards, and the flexible arm (2) is driven by means of the turning mechanism (4) to turn, so that the flexible robot is applicable to the fields of endoscopy, etc., in tubes.

Classes IPC  ?

• B25J 9/14 - Manipulateurs à commande programmée caractérisés par des moyens pour régler la position des éléments manipulateurs à fluide
• B25J 9/00 - Manipulateurs à commande programmée
• A61B 1/00 - Instruments pour procéder à l'examen médical de l'intérieur des cavités ou des conduits du corps par inspection visuelle ou photographique, p. ex. endoscopesDispositions pour l'éclairage dans ces instruments