Abrégé

An out-of-core detector calibration method, comprising: acquiring initial detection power data obtained by means of an out-of-core detector performing out-of-core detection on a target nuclear reactor, and acquiring detection power calibration parameters corresponding to the initial detection power data (S101); and using the detection power calibration parameters to perform calibration processing on the initial detection power data, so as to obtain corrected target detection power data (S102), wherein the detection power calibration parameters are calculated by means of working conditions of different water temperatures among a plurality of power working conditions in a constructed nuclear reactor, and the working conditions of different water temperatures are used for simulating the influence of different in-core water temperature distributions on the detection power data detected by means of the out-of-core detector.

Classes IPC  ?

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

Abrégé

A determination method for integrity of fuel elements in a pressurized water reactor. The determination method comprises: during operation of a pressurized water reactor, sampling a coolant of the pressurized water reactor to obtain Xe-133 specific activity and Xe-138 specific activity in the coolant sample (S10); on the basis of the Xe-133 specific activity and the Xe-138 specific activity, obtaining a fuel reliability index (S20); and on the basis of the fuel reliability index, determining whether any fuel element in the pressurized water reactor is damaged (S30). Compared with a method for determining integrity of fuel elements on the basis of specific activity of radioactive iodine isotopes, the error rate can be reduced, thereby improving the reliability of the determination result.

Classes IPC  ?

• G21C 17/02 - Dispositifs ou dispositions pour la surveillance du réfrigérant ou du modérateur
• 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

Disclosed in the present invention are a silicon carbide composite cladding and a preparation method therefor. The preparation method comprises: S1, providing a silicon carbide slurry; S2, enabling continuous silicon carbide fibers to pass through the silicon carbide slurry, so that the continuous silicon carbide fibers are impregnated with the silicon carbide slurry; S3, braiding the continuous silicon carbide fibers on the surface of an inner pipe to form a silicon carbide fiber braided layer; and S4, performing densification treatment on the silicon carbide fiber braided layer and an interface layer between the silicon carbide fiber braided layer and the inner pipe by means of chemical vapor infiltration. According to the preparation method for the silicon carbide composite cladding, the silicon carbide fiber braided layer impregnated with the silicon carbide slurry is composited on the outer surface of the inner pipe, and then densification treatment is carried out, to reduce pore defects in and between fiber bundles, improve the density, regulate the interface between the fiber and the inner pipe, and improve the toughness of the silicon carbide fiber braided layer, thereby obtaining a silicon carbide composite cladding having high strength, high thermal conductivity and high density.

Classes IPC  ?

• C04B 35/573 - Céramiques fines obtenues par frittage par réaction
• C04B 35/577 - Composites
• C04B 35/80 - Fibres, filaments, "whiskers", paillettes ou analogues
• C04B 35/622 - Procédés de mise en formeTraitement de poudres de composés inorganiques préalablement à la fabrication de produits céramiques
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages
• C04B 35/81 - "Whiskers"
• C04B 35/628 - Revêtement des poudres

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

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

A fixed related assembly of a nuclear reactor core, comprising functional rods (1) for being inserted into guide tubes of a fuel assembly (3) to achieve different functions, and a pressing system (2) for fixing and connecting the functional rods (1). The pressing system (2) comprises a central cylinder (21) having an inner cavity, and a supporting plate (22); the supporting plate (22) is fixedly connected to the central cylinder (21); the inner cavity of the central cylinder (21) comprises a first cavity (211). The pressing system (2) further comprises a pressing structure (23); the pressing structure (23) comprises a pressing rod (231), a pressing spring (232), and a spring seat (233); the pressing spring (232) fills the first cavity (211); one end of the spring seat (233) penetrates into the first cavity (211) and abuts against the lower end of the pressing spring (232), and the other end of the spring seat (233) is used for pressing against an upper tube seat of the fuel assembly (3); the pressing rod (231) is fixedly connected to the central cylinder (21), and is located at the side of the supporting plate (22) away from the spring seat (233); the pressing rod (231) is used for abutting against the side of a reactor core upper plate (4) of the nuclear reactor core close to the fuel assembly (3), so as to prevent a related assembly from axially moving.

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

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 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 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

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 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 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 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

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

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

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 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

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 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

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

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

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 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

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 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 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

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

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

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

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

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

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

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

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

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

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 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 test apparatus and method based on a coupling effect of particle irradiation and high-temperature liquid corrosion. The test apparatus comprises a particle accelerator, a first corrosive liquid housing (1), a first heating member (2), and a purging member, wherein the particle accelerator comprises an accelerator main unit (24), a beam current pipeline (25), and a transmission thin film (26). A multi-physical-field coupling environment involving a high temperature, irradiation and a corrosive liquid is formed in the test apparatus, such that a special environment is provided for the testing of an in-reactor material. In the test apparatus, mainly by means of a cooperative design of the beam current pipeline (25), the transmission thin film (26), the purging member, and the first corrosive liquid housing (1), a particle stream can be led out to an atmospheric environment; and high temperature-irradiation-corrosion coupling testing is then performed in the first corrosive liquid housing (1), such that the effect of the corrosive liquid on the accelerator main unit (24) is effectively avoided, and therefore the accelerator main unit (24) can be effectively protected, thereby improving the safety of testing.

Classes IPC  ?

• G01N 25/20 - Recherche ou analyse des matériaux par l'utilisation de moyens thermiques en recherchant la production de quantités de chaleur, c.-à-d. la calorimétrie, p. ex. en mesurant la chaleur spécifique, en mesurant la conductivité thermique
• G01N 23/00 - Recherche ou analyse des matériaux par l'utilisation de rayonnement [ondes ou particules], p. ex. rayons X ou neutrons, non couvertes par les groupes , ou

Abrégé

A driving-immersion-type sampling system for a molten metal medium in a reactor vessel (1), the system comprising a sampling pipe (10) with two open ends, a metal cable (20), a reel (30), a driving member (40) and a sampling container (50). An end of the sampling pipe (10) is used for extending below the liquid level of the molten metal medium in the reactor vessel (1); the metal cable (20) is at least partially movably provided in the sampling pipe (10) in a penetrating manner; the reel (30) and the driving member (40) are both arranged outside the sampling pipe (10); an end of the metal cable (20) is wound around the reel (30); the driving member (40) is connected to the reel (30) and is used for driving the reel (30) to rotate, so as to wind or unwind the metal cable (20); and the sampling container (50) is connected to the other end of the metal cable (20). The sampling system can allow personnel to stay away from the reactor vessel (1) during sampling, such that the risk of radioactivity during sampling is reduced, thereby improving the safety of sampling.

Classes IPC  ?

• G01N 1/10 - Dispositifs pour prélever des échantillons à l'état liquide ou fluide
• G21C 19/28 - Dispositions pour introduire un matériau fluent à l'intérieur du cœur du réacteurDispositions pour enlever un matériau fluent du cœur du réacteur

Abrégé

The present application relates to a silicon carbide cladding connection material, a silicon carbide ceramic connector and a manufacturing method therefor, and a device housing. The composition of the silicon carbide cladding connection material comprises at least five of titanium carbide, zirconium carbide, hafnium carbide, vanadium carbide, niobium carbide, tantalum carbide, chromium carbide, molybdenum carbide, and tungsten carbide.

Classes IPC  ?

• C04B 37/00 - Liaison des articles céramiques cuits avec d'autres articles céramiques cuits ou d'autres articles, par chauffage
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages

Abrégé

Disclosed is an integrated safety system for reactor, formed at a reactor pool (10) in a reactor, and relating to the technical field of nuclear power. An accommodating cavity (11) is formed on the reactor pool (10), and the integrated safety system comprises at least one safety structure. Each safety structure comprises a residual heat export loop (100) and a first pressure relief loop (200), the residual heat export loop (100) comprising a heat exchange apparatus (110), a liquid supplementing apparatus (120), a water tank (130), and a cooling apparatus (140) disposed in the water tank (130). A heat exchange water inlet of the heat exchange apparatus (110) is connected to a liquid supplementing water outlet of the liquid supplementing apparatus (120). A heat exchange water outlet of the heat exchange apparatus (110) is connected to a cooling water inlet of the cooling apparatus (140). A cooling water outlet of the cooling apparatus (140) is connected to a liquid supplementing water inlet of the liquid supplementing apparatus (120). One end of the first pressure relief loop (200) is in communication with the accommodating cavity (11), and the other end is connected to the water tank (130). The first pressure relief loop (200) is configured to be in communication with the water tank (130) when the pressure in the accommodating cavity (11) reaches a preset value.

Classes IPC  ?

• G21C 9/004 - Suppression de la pression
• 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
• 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

The present invention relates to a steam turbine protection apparatus, system and method for a pressurized water reactor. The method comprises the following steps: S1, acquiring real-time operation data of a nuclear power plant; S2, calculating real-time monitoring parameter values according to the real-time operation data of the nuclear power plant, wherein the real-time monitoring parameter values comprise a degree of superheating of steam and/or a mismatch amount of a water supply flow with respect to the nuclear power of a reactor core; S3, comparing the real-time monitoring parameter values with fixed protection values, so as to determine whether there is a risk during the operation of a steam turbine, and outputting a determination result, and according to the determination result, controlling the steam turbine to execute an action. The apparatus comprises a processing module and a control module. The system comprises a reactor core, a steam turbine, a direct-current steam generator, a condenser, a heat transfer pipeline, a water supply pipeline and the steam turbine protection apparatus. By means of the present invention, the state of a steam turbine can be monitored in real time, protection regarding the situation where the degree of superheating of steam is low is implemented for the steam turbine, and/or protection regarding a situation where the mismatch amount of a water supply flow with respect to the nuclear power of a reactor core is great is implemented for the steam turbine, such that the steam turbine is triggered to trip, thereby preventing the steam from entering the steam turbine.

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
• F01D 21/00 - Arrêt des "machines" ou machines motrices, p. ex. dispositifs d'urgenceDispositifs de régulation, de commande ou de sécurité non prévus ailleurs
• G21D 3/06 - Dispositions de sécurité réagissant à des défaillances à l'intérieur de l'installation

Abrégé

A double-loop small pressurized water reactor emergency residual heat removal system, comprising a first loop (10) and a second loop (20), each of which comprises a steam generator (11, 21), a main steam pipeline (12, 22), a main steam isolation valve (13, 23), a condenser (14, 24), a water conveying tank (15, 25), a condensate water suction pump (17, 27), a check valve (18, 28), a main water supply isolation valve (19, 29), a main water supply check valve (110, 210), and a main water supply pipeline (111, 211). The emergency residual heat removal system of the first steam generator (11) comprises the first loop (10) and a first redundant loop which comprises a second steam communication pipeline (212), a second steam communication isolation valve (213), a second condenser (24), a second water conveying tank (25), a second condensate water suction pump (27), a second water supply communication isolation valve (214), a second water supply communication check valve (215), and a second water supply communication pipeline (216); and the emergency residual heat removal system of the second steam generator comprises the second loop (20) and a second redundant loop which comprises a first steam communication pipeline (112), a first steam communication isolation valve (113), a first condenser (14), a first water conveying tank (15), a first condensate water suction pump (17), a first water supply communication isolation valve (114), a first water supply communication check valve (115), and a first water supply communication pipeline (116).

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur
• G21D 1/02 - Dispositions d'équipement auxiliaire

Abrégé

The present application relates to a reactor core control method and apparatus (400), a computer device, a storage medium, and a program product. The method comprises: controlling a first rod group, a second rod group, and a third rod group to act according to an average temperature deviation between an actual temperature value and a set temperature value of a coolant of a reactor core, a rod position deviation between a target rod position and an actual rod position of the first rod group, and an axial power deviation difference value between an actual axial power deviation and a preset axial power deviation of the reactor core.

Classes IPC  ?

• G06Q 10/06 - Ressources, gestion de tâches, des ressources humaines ou de projetsPlanification d’entreprise ou d’organisationModélisation d’entreprise ou d’organisation
• G21D 3/16 - Régulation de différents paramètres dans l'installation par ajustement du réacteur en réponse uniquement aux changements se produisant dans la demande du moteur en variant la réactivité
• G21D 3/00 - Commande des installations à énergie nucléaire
• G21C 7/08 - Commande de la réaction nucléaire par application de matériau absorbant les neutrons, c.-à-d. matériau avec section efficace d'absorption excédant largement la section efficace de réflexion par déplacement des éléments de commande solides, p. ex. barres de commandes

Abrégé

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

Classes IPC  ?

• G21D 3/04 - Dispositions de sécurité

Abrégé

An autonomous control method and apparatus of a nuclear reactor, a computer device, a storage medium, and a computer program product. The method comprises: determining a target state trajectory and a reactor full-state trajectory of a nuclear reactor (102); performing trajectory optimization on the reactor full-state trajectory according to the target state trajectory to obtain a control action combination of the nuclear reactor (104), the control action combination being used for ensuring that the deviation between the target state trajectory and the reactor full-state trajectory reaches a preset value and meets the dynamic evolution characteristics of the reactor; and performing autonomous control on the nuclear reactor on the basis of the control action combination (106). The use of the method improves the control precision of the nuclear reactor.

Classes IPC  ?

• G05B 13/04 - Systèmes de commande adaptatifs, c.-à-d. systèmes se réglant eux-mêmes automatiquement pour obtenir un rendement optimal suivant un critère prédéterminé électriques impliquant l'usage de modèles ou de simulateurs

Abrégé

The present application relates to a method for calculating creep deformation of a finite-length cladding tube, comprising: acquiring structural parameters of a finite-length cladding tube (102); performing discrete processing on the structural parameters, and obtaining a finite element mesh model corresponding to the finite-length cladding tube (104); calculating stiffness information and load information according to model parameters of the finite element mesh model and the structure parameters (106); and calculating creep deformation results of the finite-length cladding tube according to the stiffness information, the load information, a preset additional load vector, and a preset creep deformation calculation relationship (108).

Classes IPC  ?

• G06F 30/17 - Conception mécanique paramétrique ou variationnelle
• 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 111/10 - Modélisation numérique
• G06F 119/02 - Analyse de fiabilité ou optimisation de fiabilitéAnalyse de défaillance, p. ex. performance dans le pire scénario, analyse du mode de défaillance et de ses effets [FMEA]
• G06F 119/14 - Analyse des forces ou optimisation des forces, p. ex. forces statiques ou dynamiques

Abrégé

A control rod and a control rod assembly. The control rod comprises a core block (100), a bushing assembly (200), and a cladding (300); the bushing assembly (200) comprises a first bushing (210) sleeved on the core block (100); the cladding (300) is sleeved on the first bushing (210); the first bushing (210) is configured to form a constraint structure surrounding the core block (100) to limit the expansion of the core block (100) in a radial direction of the core block (100).

Classes IPC  ?

• G21C 7/10 - Structures des éléments de commande

Abrégé

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

Classes IPC  ?

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

Abrégé

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

Classes IPC  ?

• G21C 17/108 - Mesure du flux

Abrégé

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

Classes IPC  ?

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

Abrégé

A high-uranium-density ATF nuclear fuel pellet and a preparation method therefor. The high-uranium-density ATF nuclear fuel pellet comprises a columnar inner region (10), an isolation layer (20) covering the outer surface of the inner region (10), and a peripheral region (30) covering the isolation layer (20), wherein the inner region (10) is prepared from a nuclear fuel that is resistant to the oxidation of water and water vapor, and a high-uranium-density nuclear fuel; the peripheral area (30) is prepared from a nuclear fuel that is resistant to the oxidation of water and water vapor; and the mass of the high-uranium-density nuclear fuel accounts for 3-50% of the total mass of the inner region (10) and the isolation layer (20). The high-uranium-density ATF nuclear fuel pellet has an increased uranium loading amount and an effectively reduced central temperature, thereby improving the economical efficiency and safety of the operation of a reactor; and the use of the high-uranium-density nuclear fuel in a water-cooled nuclear reactor is realized, and the accident-tolerant capability of the nuclear reactor is improved.

Classes IPC  ?

• G21C 3/62 - Combustible céramique
• G21C 21/10 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives par extrusion, filage, étirage

Abrégé

According to the nuclear fuel transportation container provided by the present invention, a tank body for accommodating a nuclear fuel assembly is wrapped in a plastically deformable impact-resistant protective layer, and several recesses for accommodating supporting columns are formed in the protective layer, so that the supporting columns can abut against the tank body and bear the weight of the container; the protective layer does not bear weight, and an energy absorption structure is used to plastically deform and resist impact, so that the impact resistance is improved.

Classes IPC  ?

• G21F 5/08 - Amortisseurs de chocs spécialement adaptés à ces récipients
• G21F 5/06 - Détails ou accessoires des récipients

Abrégé

A nuclear reactor design scheme optimization method and apparatus, a computer device, a storage medium, and a computer program product. The method comprises: acquiring an overall machine learning model of a nuclear reactor; according to the overall machine learning model of the nuclear reactor, searching for a nuclear reactor design scheme using an optimization algorithm, to obtain a nuclear reactor design scheme search result; performing validation and regression of the nuclear reactor design scheme search result; and generating an optimized nuclear reactor design scheme according to a validation and regression result. In the whole process, diversified nuclear reactor design schemes are generated using a machine learning model, and an optimization algorithm is used to perform a search of the schemes, and to perform subsequent validation and regression, so that an optimized nuclear reactor design scheme can be generated efficiently while ensuring the accuracy of scheme construction.

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

Abrégé

A control rod for nuclear fuel and a cladding thereof. The cladding of the control rod comprises a first cladding tube (10), a second cladding tube (20) and a bidirectional end plug (30) connected therebetween. A gas storage chamber (100) is formed in the first cladding tube (10), and an accommodating chamber (200) for holding an absorber (40) is formed in the second cladding tube (20). A through hole (300) penetrating through two opposite ends of the bidirectional end plug (30) is formed therein, the through hole (300) making the gas storage chamber (100) and the accommodating chamber (200) communicate. The cladding of the control rod comprises the separate parts which are connected into a whole by means of the bidirectional end plug (30). The first cladding tube (10) and the second cladding tube (20), as the separate parts of the cladding, are respectively used for forming the gas storage chamber (100) and holding the absorber (40). The gas storage chamber (100) is independent of the inner space in which the absorber (40) is placed and effectively stores a large amount of gas released when the absorber (40) absorbs neutrons, thereby avoiding expansion of the cladding tube in which the absorber (40) is arranged, and prolonging the service life of the cladding and the control rod.

Classes IPC  ?

• G21C 7/10 - Structures des éléments de commande
• G21C 7/103 - Ensembles de commande comportant un ou plusieurs absorbants ainsi que d'autres éléments, p. ex. combustibles ou modérateurs

Abrégé

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

Classes IPC  ?

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

Abrégé

An on-line detection device (1) and method for impurities in a lead-bismuth coolant (3). The on-line detection device (1) for impurities in the lead-bismuth coolant (3) comprises a measurement channel (10), a pulse laser (20), an optical lens (30), and a spectrometer (40); the measurement channel (10) is connected between the top cover of a pressure vessel (2) and the lead-bismuth coolant (3) inside the pressure vessel (2); the pulse laser (20) and the spectrometer (40) are both provided above the top cover of the pressure container (2); the optical lens (30) is arranged in the measurement channel (10) and is located on an emitting light path of the pulse laser (20), and focuses a pulse laser beam emitted by the pulse laser (20) and refracts same to the surface of the lead-bismuth coolant (3) to generate plasma; the spectrometer (40) is connected to the measurement channel (10), and acquires the plasma to obtain the spectral line intensities of impurity elements. By means of cooperation of the measurement channel (10), the pulse laser (20), and the spectrometer (40), the content of impurities in the lead-bismuth coolant (3) can be accurately detected by LIBS, the urgent need for multi-element, in-situ, and remote online measurement and analysis of impurities in a lead-based alloy is met, and the online monitoring requirement for reactor operation is met.

Classes IPC  ?

• G01N 21/63 - Systèmes dans lesquels le matériau analysé est excité de façon à ce qu'il émette de la lumière ou qu'il produise un changement de la longueur d'onde de la lumière incidente excité optiquement
• G21C 17/025 - Dispositifs ou dispositions pour la surveillance du réfrigérant ou du modérateur pour la surveillance de réfrigérants ou de modérateurs liquides pour la surveillance de réfrigérants constitués par des métaux liquides

Abrégé

Disclosed in the present invention is a fuel assembly locking device, comprising a locking piece disposed at the lower end of a fuel assembly and elastically engaged in a mounting hole of a lower core plate, and an anti-loosening mechanism disposed below the lower core plate and mated with the locking piece. The anti-loosening mechanism comprises a guide tube and an anti-loosening rod; the anti-loosening rod is disposed in the guide tube and moves back and forth between a first position and a second position in the axial direction of the guide tube; at the first position, an end part of the anti-loosening rod is ejected out of the guide pipe and mated in the locking piece, to prevent the locking piece from being disengaged from the mounting hole; at the second position, the end part of the anti-loosening rod is separated from the locking piece and retracted in the guide tube, so that the restriction on the locking piece is relieved. According to the fuel assembly locking device according to the present invention, the fuel assembly is fixed on the lower core plate by means of mating between the locking piece and the anti-loosening mechanism, the structure and operation are simple, and compared with an existing rotating mode, compact arrangement of fuel assemblies is facilitated.

Classes IPC  ?

• G21C 3/12 - Moyens formant partie de l'élément pour le positionner dans le cœur du réacteurEntretoises extérieures à cet effet
• G21C 5/06 - Moyens pour placer ou pour supporter les éléments combustibles
• G21C 3/14 - Moyens formant partie de l'élément pour le placer ou le retirer du cœurMoyens pour assembler des éléments adjacents

Abrégé

A nuclear reaction device and a purification system. The nuclear reaction device comprises: a reaction vessel (100) provided with a reaction chamber (110) and a top cover (120) for sealing the reaction chamber (110); a reaction module (200) arranged in the reaction chamber (110), wherein the reaction module (200) reacts in the reaction chamber (110) and releases heat; a heat exchange module (300) arranged in the reaction chamber (110), wherein the heat exchange module (300) can absorb the heat released by the reaction module (200), an inlet of the heat exchange module (300) is configured to be in communication with a liquid supply system, and an outlet of the heat exchange module (300) is configured to be in communication with a thermally driven system.

Classes IPC  ?

• G21C 15/18 - Dispositions pour le refroidissement d'urgenceMise hors circuit de la chaleur
• G21C 15/12 - 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 de l'enceinte sous pressionAménagement 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 de l'enceinte d'enveloppe
• G21C 19/30 - Dispositions pour introduire un matériau fluent à l'intérieur du cœur du réacteurDispositions pour enlever un matériau fluent du cœur du réacteur avec purification continue du matériau fluent en circulation, p. ex. par extraction des produits de fission

Abrégé

2322 hybrid nanofiber membrane. The present invention has a low preparation cost and simple operation; the prepared inorganic nanofiber membrane can be used to efficiently filter an aerosol under high-temperature conditions; and particularly, the filtering efficiency for a polonium aerosol can reach 99%.

Classes IPC  ?

• B01D 67/00 - Procédés spécialement adaptés à la fabrication de membranes semi-perméables destinées aux procédés ou aux appareils de séparation
• B01D 71/02 - Matériaux inorganiques
• B01D 46/54 - Séparateurs de particules utilisant des feuilles ou diaphragmes filtrants à structure ultra-fine, p. ex. appareils de précipitation de poussières
• D04H 1/728 - Non-tissés formés uniquement ou principalement de fibres coupées ou autres fibres similaires relativement courtes caractérisés par la méthode de formation des voiles ou couches, p. ex. par la réorientation des fibres les fibres étant disposées au hasard par électrofilage
• D01D 5/00 - Formation des filaments, fils ou similaires
• D01F 9/08 - Filaments, ou similaires, faits par l’homme, formés d’autres substancesLeur fabricationAppareils spécialement adaptés à la fabrication de filaments de carbone de matière inorganique
• D01F 8/18 - Filaments, ou similaires, faits par l’homme, conjugués, c.-à-d. à plusieurs composantsLeur fabrication à partir d'autres substances

Abrégé

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

Classes IPC  ?

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

Abrégé

Disclosed in the present invention are a fuel rod, a fuel assembly, and a reactor core. The fuel rod comprises a cladding and a fuel pellet sealed within the cladding, wherein the fuel pellet comprises a base portion arranged in the center and at least six wing portions extending outward from the base portion, each wing portion being circumferentially twisted at least one turn with a longitudinal axis of the base portion as the center. The fuel assembly comprises an outer box and the fuel rods, wherein the outer box has a hexagonal cross-section, and the fuel rods are in a triangular arrangement in the outer box. The reactor core comprises the fuel assembly described above. According to the present invention, the fuel rod has at least six wing portions, which are each circumferentially twisted at least one turn with the longitudinal axis of the base portion as the center, and thus has more excellent thermal conductivity and can easily remove the internal heat of the fuel rod to greatly reduce the central temperature of the fuel pellet, and the reactor core has the potential to increase the power density under the condition of ensuring nuclear safety, so that a reactor is safer and more economical, and has a compact structure and a wide application range.

Classes IPC  ?

• G21C 3/28 - Éléments combustibles avec un matériau fissile ou surrégénérateur sous forme solide à l'intérieur d'une enveloppe non active
• G21C 3/322 - Moyens pour influencer l'écoulement du réfrigérant à travers ou autour des faisceaux
• G21C 3/16 - Détails de structure à l'intérieur de l'enveloppe
• G21C 3/326 - Faisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant des éléments combustibles de différentes compositionsFaisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles comprenant, en plus des éléments combustibles, d'autres éléments en forme d'aiguille, de barre ou de tube, p. ex. barres de commande, barres de support de grilles, barres fertiles, barres à poison ou barres factices

Abrégé

Disclosed in the present invention is a reactor primary circuit main device installation method. The method comprises the following steps: S1) pre-assembling a reactor pressure container, in-reactor members, an inner sleeve, a steam generator and a main pump case into a container module, and pre-assembling a top cover of the reactor pressure container, a control rod driving mechanism and a reactor top structure into a reactor top module; S2) installing a main device support and a voltage stabilizer support; S3) hoisting the container module onto the main device support, adjusting the position of the container module, and installing the container module on the main device support; and S4) installing a voltage stabilizer on the voltage stabilizer support, connecting a main pump electric motor and hydraulic components to the main pump case, and installing the reactor top module on the container module. Compared with the prior art, according to the reactor primary circuit main device installation method of the present invention, various devices are pre-assembled into the container module and the reactor top module in a device final-assembly plant, and the container module and the reactor top module are then transported to an installation site for installation, such that the installation precision and efficiency can be improved.

Classes IPC  ?

• G21D 1/00 - Détails des installations à énergie nucléaire

Abrégé

A pressure suppression pool apparatus and a nuclear reactor. The pressure suppression pool apparatus comprises a housing (1) and at least one hollow pressure suppression pipe (2); a closed cavity and a plurality of baffles (3) for slowing gas flow are provided in the housing (1); the cavity is filled with a cooling liquid (11); the baffles (3) are submerged in the cooling liquid (11); an accommodating space (12) for accommodating gas is provided on the cooling liquid (11) inside the cavity; the pressure suppression pipe (2) passes through the outer wall of the housing (1), and comprises an out-of-housing section (21) located outside the housing (1) and an in-housing section (22) located in the housing (1); an end portion of the in-housing section (22) is inserted into the cooling liquid (11), and the out-of-housing section (21) is connected to the gas in an environment. The present pressure suppression pool apparatus is a passive system, has high safety and reliability, high cooling efficiency, and a significant pressure suppression effect, can effectively suppress the rise of the pressure and temperature of a containment under predetermined working conditions, and has low volume requirements for the space (12) and the cooling liquid (11).

Classes IPC  ?

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

Abrégé

Disclosed are a plate-type fuel assembly and a reactor core. The plate-type fuel assembly comprises a metal cylinder and multiple groups of fuel structures that are provided within the metal cylinder; the multiple groups of the fuel structures are sequentially stacked within the metal cylinder and separately extend in the length direction of the metal cylinder; and each group of the fuel structures comprises a fuel layer and a cooling layer that is stacked on at least one side of the fuel layer; and the cooling layer is provided with multiple coolant channels arranged at intervals, and the coolant channels penetrate two opposite end surfaces of the cooling layer and respectively communicate with two open ends of the metal cylinder. The plate-type fuel assembly of the present invention is used for forming a reactor core, has high structural strength due to being formed by stacking the fuel layers and cooling layers within the metal cylinder, an increased heat transfer area due to the provision of multiple coolant channels of the cooling layer, increases the ratio of heat transfer area to volume, can improve the heat transfer coefficient and heat transfer density of a reactor core, and is suitable for the construction of small reactor cores and high-temperature reactors.

Classes IPC  ?

• G21C 3/32 - Faisceaux d'éléments combustibles en forme d'aiguilles, de barres ou de tubes parallèles
• G21C 15/00 - Dispositions pour le refroidissement à l'intérieur de l'enceinte sous pression contenant le cœurEmploi de réfrigérants spécifiques

Abrégé

Disclosed in the present invention are a fuel module and an application thereof. The fuel module comprises a column structure, a plurality of hollow closed holes and a plurality of cooling through holes that are dispersedly provided in the column structure, and fuel provided in the column structure; the hollow closed holes are located in the column structure; each of the hollow closed holes is filled with fuel; the cooling through holes are isolated from the hollow closed holes, run through two opposite sides of the column structure, and are used for allowing a cooling medium to pass through. In the fuel module of the present invention, a column structure is used as a supporting body, and hollow closed holes are filled with fuel and cooperate with cooling through holes to form an integrated structure. Compared with the cladding-pellet-assembly structure of an existing fuel assembly, the overall structure has high strength, and the fuel module can withstand high pressure, is safe and reliable, and improves the safety of a nuclear reactor.

Classes IPC  ?

• G21C 3/04 - Détails de structure
• G21C 3/16 - Détails de structure à l'intérieur de l'enveloppe

Abrégé

A silicon carbide joint and a metal penetration connection method therefor. The metal penetration connection method comprises the following steps: S1. formulating a first metal raw material and a second metal raw material to form a connection material; S2. arranging the connection material between the connection faces of two silicon carbide parts, so that the connection material and the two silicon carbide parts form a sandwich structure; S3. placing the sandwich structure in a sintering environment having a metal phase and sintering same, wherein during the sintering process, the metal phase penetrates into the connection material and fills air holes generated in the connection material; and S4. after sintering, densifying the connection material to form a connection layer, so as to connect the two silicon carbide parts to form a silicon carbide joint. According to the metal penetration connection method for the silicon carbide joint, the metal phase penetrates into the connection material, so that the densification of the connection material is achieved, the silicon carbide joint with high strength connection and good air tightness is obtained, and the requirements for the shape, structure, etc., of silicon carbide are reduced.

Classes IPC  ?

• C04B 37/00 - Liaison des articles céramiques cuits avec d'autres articles céramiques cuits ou d'autres articles, par chauffage

Abrégé

A control rod position detection system and a device. The control rod position detection system comprises rod position detectors (10), power supply processing apparatuses (20), and coding processing apparatuses (30). Each rod position detector (10), a coding processing apparatus (30), and a power supply processing apparatus (20) form a measurement channel. The power supply processing apparatuses (20) each perform phase processing on a power supply signal after receiving the power supply signal, to obtain a specific current signal which cancels out a current interference signal of another rod position detector (10) around any rod position detector (10), and transmits the current signal to the corresponding rod position detector (10). The control rod position detection device comprises rod position detectors (10), power supply processing apparatuses (20), coding processing apparatuses (30), and two or more measurement cabinets (100). Each rod position detector (10) generates a sensing voltage according to a corresponding current signal, and finally a coding processing apparatus (30) processes the sensing voltage to obtain a position-encoding signal representing position information of a control rod.

Classes IPC  ?

• G21C 17/10 - Combinaison structurelle de l'élément combustible, de la barre de commande, du cœur du réacteur, ou de la structure du modérateur avec des instruments sensibles, p. ex. pour la mesure de la radioactivité, des contraintes
• G01D 5/14 - Moyens mécaniques pour le transfert de la grandeur de sortie d'un organe sensibleMoyens pour convertir la grandeur de sortie d'un organe sensible en une autre variable, lorsque la forme ou la nature de l'organe sensible n'imposent pas un moyen de conversion déterminéTransducteurs non spécialement adaptés à une variable particulière utilisant des moyens électriques ou magnétiques influençant la valeur d'un courant ou d'une tension

Abrégé

Disclosed in the present invention are a high-entropy ceramic inert matrix dispersion fuel pellet and a preparation method therefor. The preparation method comprises: S1, preparing raw material powder into a slurry and mixed powder, respectively; S2, spraying the slurry on the surfaces of rolling fuel particles, and drying to form a coating layer adhered to the surface of the fuel particles; S3, preparing a core biscuit and a core-shell biscuit; S4, loading the core biscuit into the core-shell biscuit, and performing heat treatment to obtain a ceramic green body; and S5, performing solid-melt sintering to form a densified high-entropy ceramic inert matrix dispersion fuel pellet. The high-entropy ceramic inert matrix dispersion fuel pellet is prepared by means of an in-situ reaction-solid solution sintering two-step method, and the characteristics that the fuel pellet is high in heat conductivity, low in swelling, easy to post-process, suitable for industrial production and the like are achieved; the nuclear fuel containment under the condition that a nuclear reactor cladding is damaged is improved, the leakage of the nuclear fuel is prevented, and the nuclear power safety is improved; and the high-temperature resistance of the nuclear fuel is improved, and the application of the nuclear fuel in a high-temperature reactor is promoted.

Classes IPC  ?

• C04B 35/56 - 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
• C04B 35/58 - 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 borures, nitrures ou siliciures
• C04B 35/622 - Procédés de mise en formeTraitement de poudres de composés inorganiques préalablement à la fabrication de produits céramiques
• G21C 3/62 - Combustible céramique
• G21C 3/64 - Combustible céramique en dispersion, p. ex. cermet
• G21C 21/02 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives

Abrégé

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

Classes IPC  ?

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

Abrégé

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

Classes IPC  ?

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

Abrégé

A reactor starting method and system. The method comprises the following steps: S1, when a core (3) of a reactor is started, controlling a primary loop main pump (1) of the reactor to be started at a first rotation speed; S2, during the process of the power of the core (3) rising from zero power to full power, acquiring real-time power of the core (3); S3, determining whether the real-time power of the core (3) is within a preset dual-speed switching power range; and S4, if so, controlling the main pump (1) to be switched from the first rotation speed to a second rotation speed. By means of the method, the aim of saving power can be achieved, a pump starting current is low, the starting characteristics are better, and the reserve power of an electric motor can be greatly reduced, thereby facilitating the reduction of the size of the electric motor.

Classes IPC  ?

• G21C 7/32 - Commande de la réaction nucléaire par variation du courant de réfrigérant à travers le cœur
• G21D 3/14 - Régulation de différents paramètres dans l'installation par ajustement du réacteur en réponse uniquement aux changements se produisant dans la demande du moteur en variant le flux du réfrigérant

Abrégé

A high-entropy ceramic inert matrix dispersion fuel pellet and a preparation method therefor. The high-entropy ceramic inert matrix dispersion fuel pellet comprises a cylindrical fuel-free area (20) and a column-shaped fuel area (10) provided in the fuel-free area (20); the fuel area (10) comprises a high-entropy ceramic matrix (11) and fuel particles (12) dispersed in the high-entropy ceramic matrix (11); the high-entropy ceramic matrix (11) and the fuel-free area (20) are prepared by sintering high-entropy ceramic. The high-entropy ceramic is used as a matrix of the inert matrix dispersion fuel pellet, such that the pellet is high in thermal conductivity, low in swelling rate, easy to post-process, and suitable for industrial production.

Classes IPC  ?

• G21C 3/64 - Combustible céramique en dispersion, p. ex. cermet
• G21C 21/02 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives
• G21C 21/10 - Fabrication des éléments combustibles ou surrégénérateurs à l'intérieur de gaines non-actives par extrusion, filage, étirage

Abrégé

An induction heating-based silicon carbide cladding connecting method and a silicon carbide cladding. The induction heating-based silicon carbide cladding connecting method comprises: adding a connecting material and a dispersing agent into an organic solvent for dispersing to form a mixed slurry; evenly spreading the mixed slurry on connecting surfaces of an SiC end plug and/or an SiC cladding tube, and aligning the connecting surfaces to match the SiC end plug and the SiC cladding tube to form a connecting structure; curing the connecting structure under a protective atmosphere, the mixed slurry being cured to form a connecting layer; placing the connecting structure on an induction heating device for induction heating to densify the connecting layer, so as to densely connecting the SiC end plug to the SiC cladding tube to form a SiC cladding. In the present invention, the use of induction heating to realize quick connection between the end plug and the cladding tube greatly saves working time and improves connecting efficiency, and conductive treatment on the high-purity non-conductive SiC cladding tube and end plug is unnecessary in induction heating, such that only the connecting layer is heated, thereby achieving high reliability.

Classes IPC  ?

• C04B 37/00 - Liaison des articles céramiques cuits avec d'autres articles céramiques cuits ou d'autres articles, par chauffage
• G21C 21/00 - Appareillage ou procédés spécialement adaptés pour la fabrication des réacteurs ou de pièces de ceux-ci
• G21C 3/10 - Obturateurs d'extrémités

Abrégé

A reactor primary circuit spray system, comprising: a steam-powered spray pump (40), a water inlet of the steam-powered spray pump (40) being used for being connected to a third pipeline (704) on a primary circuit, and a water outlet of the steam-powered spray pump (40) being used for being connected to a voltage regulator (50) on the primary circuit; and a steam pipeline (701), one end of the steam pipeline (701) being used for being connected to a steam outlet of a steam generator (20) on the primary circuit, and the other end of the steam pipeline (701) being used for being connected to a pressure head of the steam-powered spray pump (40).

Classes IPC  ?

• G21C 15/25 - Cyclage du fluide réfrigérant pour des liquides utilisant des pompes à jet

Abrégé

Disclosed are a zirconium alloy and a manufacturing method therefor, a cladding tube, and a fuel assembly, wherein the zirconium alloy comprises the following components in percentages by mass: 1.20% to 1.40% of niobium, 0.03% to 0.07% of vanadium, 0.12% to 0.15% of oxygen, and the balance of zirconium. The zirconium alloy of the present invention has excellent corrosion resistance, can significantly reduce hydrogen absorption amount and improve high-temperature oxidation quenching resistance, has a more excellent corrosion resistance, a better hydrogen absorption resistance and a better embrittlement resistance after high-temperature oxidation quenching compared with an existing Zr-4 alloy, and is suitable for a fuel assembly of a nuclear power plant reactor and improves the service performance and safety performance of the fuel assembly as a cladding material of the fuel assembly.

Classes IPC  ?

• C22C 16/00 - Alliages à base de zirconium
• C22C 1/02 - Fabrication des alliages non ferreux par fusion
• C22F 1/18 - Métaux réfractaires ou à point de fusion élevé ou leurs alliages
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages

Abrégé

Disclosed are a zirconium alloy and a fabrication method therefor, a cladding tube, and a fuel assembly, the zirconium alloy comprising the following components by mass percentage: 0.45%-0.95% of niobium, 0.21%-0.35% of tin, 0.03%-0.1% of iron, 0.03%-0.1% of vanadium, and 1000 ppm-1600 ppm of oxygen and the remainder is zirconium, wherein the total amount of iron and vanadium is ≤0.15%. The zirconium alloy of the present invention has excellent corrosion resistance and good embrittlement resistance after high-temperature oxidation and quenching by means of the ratio of the components, is suitable for cladding material of nuclear power plant reactors, and improves the service performance and safety of a fuel assembly.

Classes IPC  ?

• C22C 16/00 - Alliages à base de zirconium
• C22C 1/02 - Fabrication des alliages non ferreux par fusion
• C22F 1/18 - Métaux réfractaires ou à point de fusion élevé ou leurs alliages
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages

Abrégé

Disclosed are a zirconium alloy and a preparation method therefor, a cladding tube, and a fuel assembly. The zirconium alloy comprises the following components in percentages by mass: 0.48% to 0.95% of niobium, 0.37% to 0.75% of tin, 0.03% to 0.15% of iron, and 0 to 0.15% of vanadium, and further comprises 1100 ppm to 1600 ppm of oxygen and the balance of Zr, wherein the contents of niobium, iron and vanadium satisfy the following relationship: (Nb-0.45％)≥Fe+V and Fe+V≤0.2%. Due to the proportion of each component, the zirconium alloy of the present invention has a more excellent corrosion resistance and creep resistance compared with an existing Zr-4 alloy, and is suitable for cladding, grillwork and a guide tube of a fuel assembly of a nuclear power plant reactor and improves the service performance and safety performance of the fuel assembly.

Classes IPC  ?

• C22C 16/00 - Alliages à base de zirconium
• C22C 1/03 - Fabrication des alliages non ferreux par fusion utilisant des alliages-mères
• C22F 1/18 - Métaux réfractaires ou à point de fusion élevé ou leurs alliages
• C21D 9/52 - Traitement thermique, p. ex. recuit, durcissement, trempe ou revenu, adapté à des objets particuliersFours à cet effet pour fils métalliquesTraitement thermique, p. ex. recuit, durcissement, trempe ou revenu, adapté à des objets particuliersFours à cet effet pour bandes métalliques
• C21D 9/08 - Traitement thermique, p. ex. recuit, durcissement, trempe ou revenu, adapté à des objets particuliersFours à cet effet pour corps tubulaires ou tuyaux
• B21J 5/00 - Méthodes pour forger, marteler ou presserÉquipement ou accessoires particuliers
• G21C 3/07 - EnveloppesChemises caractérisées par le matériau, p. ex. alliages

Abrégé

Disclosed in the present invention are a connection method for a silicon carbide (SiC) cladding for nuclear application, and a SiC cladding and an application thereof. The connection method comprises the following steps: S1, preparing a connection slurry; S2, uniformly applying the connection slurry on connection surfaces of a SiC end plug and/or a SiC cladding tube, and relatively matching the connection surfaces of the SiC end plug and the SiC cladding tube to form a connection structure; S3, heating the connection structure to 100-300°C in a protective atmosphere, keeping the temperature for 0.1-4 h, and curing the connection slurry between the SiC end plug and the SiC cladding tube to form a connection layer; and S4, performing resistance welding treatment on the cured connection structure to densify the connection layer, and connecting the SiC end plug and the SiC cladding tube densely to form the SiC cladding. According to the connection method for a SiC cladding for nuclear application in the present invention, rapid connection between the end plug and the cladding tube is achieved by using the resistance welding technology, thereby greatly saving the working time, and improving the connection efficiency; and a heat-affected zone of resistance welding connection is small, such that a nuclear fuel inside the cladding cannot be affected, and the reliability of the cladding is improved.

Classes IPC  ?

• C04B 37/00 - Liaison des articles céramiques cuits avec d'autres articles céramiques cuits ou d'autres articles, par chauffage
• 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

Abrégé

A control rod absorber assembly (1), a control rod (3) and a control rod assembly. The control rod absorber assembly (1) comprises a first absorber (11), a second absorber (12) and a stress transmission assembly (13), wherein the stress transmission assembly (13) is arranged around the periphery of the second absorber (12) and forms a stress combination together with the second absorber (12), and a receding space (2) for accommodating the second absorber (12) when the second absorber expands in a radial direction is provided between the stress transmission assembly (13) and the periphery of the second absorber (12); and the first absorber (11) and the stress combination are configured to be mounted in a cladding tube (31) of a control rod (3), the first absorber (11) and the stress combination are arranged in an axial direction of the cladding tube (31), and the first absorber (11) is located above the stress combination and is configured for pressing the stress combination downwards. By arranging the stress transmission assembly (13), the second absorber (12) does not bear the action of gravity of the first absorber (11), so as to reduce compression creep of the second absorber (12); and when the second absorber (12) absorbs neutrons and expands in the radial direction, an expanded portion can be accommodated in a receding space (2), so as to prevent a cladding tube (31) from being damaged.

Classes IPC  ?

• G21C 7/103 - Ensembles de commande comportant un ou plusieurs absorbants ainsi que d'autres éléments, p. ex. combustibles ou modérateurs
• G21C 7/10 - Structures des éléments de commande
• G21C 7/11 - Éléments de commande déformables, p. ex. flexibles, télescopiques, articulés

Abrégé

The present disclosure relates to a method and apparatus for controlling a pressurized water reactor core, and a computer device and a storage medium. The method comprises: determining an operation mode of a pressurized water reactor core, wherein the pressurized water reactor core is loaded with at least one first control rod group, a second control rod group, a third control rod group and a plurality of fuel assemblies; acquiring a rod group control instruction corresponding to the operation mode; and according to the rod group control instruction, controlling the pressurized water reactor core on the basis of the first control rod group, the second control rod group and the third control rod group. By using the method, the aim of automatically controlling a pressurized water reactor core by means of a first control rod group, a second control rod group and a third control rod group is realized, and it is not necessary to control the pressurized water reactor core by frequently adjusting the concentration of boron in a primary circuit coolant, such that the control flow of the pressurized water reactor core is simplified, and thus the control efficiency of the pressurized water reactor core is improved.

Classes IPC  ?

• G21D 3/16 - Régulation de différents paramètres dans l'installation par ajustement du réacteur en réponse uniquement aux changements se produisant dans la demande du moteur en variant la réactivité
• G21D 3/00 - Commande des installations à énergie nucléaire
• 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