Abstract

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).

IPC Classes  ?

• G21C 17/00 - MonitoringTesting

Abstract

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.

IPC Classes  ?

• G21D 3/08 - Regulation of any parameters in the plant
• H02J 3/24 - Arrangements for preventing or reducing oscillations of power in networks
• G21D 3/00 - Control of nuclear power plant
• G21D 3/10 - Regulation of any parameters in the plant by a combination of a variable derived from neutron flux with other controlling variables, e.g. derived from temperature, cooling flow, pressure
• G05B 19/04 - Programme control other than numerical control, i.e. in sequence controllers or logic controllers

Abstract

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).

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G21C 17/108 - Measuring reactor flux
• G21C 17/06 - Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination
• G06F 17/18 - Complex mathematical operations for evaluating statistical data

Abstract

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.

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G06F 18/2433 - Single-class perspective, e.g. one-against-all classificationNovelty detectionOutlier detection

Abstract

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).

IPC Classes  ?

• G21D 3/06 - Safety arrangements responsive to faults within the plant
• G06F 18/24 - Classification techniques

Abstract

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.

IPC Classes  ?

• G21C 5/14 - Moderator or core structureSelection of materials for use as moderator characterised by shape

Abstract

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).

IPC Classes  ?

• G21C 17/108 - Measuring reactor flux

Abstract

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.

IPC Classes  ?

• G21C 17/108 - Measuring reactor flux

Abstract

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.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation
• G21C 17/00 - MonitoringTesting

Abstract

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.

IPC Classes  ?

• F22B 35/00 - Control systems for steam boilers
• F22B 37/46 - Applications, arrangements or dispositions of alarm or automatic safety devices responsive to low or high water level, e.g. for checking, suppressing or extinguishing combustion in boilers
• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

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.

IPC Classes  ?

• G01N 25/16 - Investigating or analysing materials by the use of thermal means by investigating thermal coefficient of expansion
• G01N 25/18 - Investigating or analysing materials by the use of thermal means by investigating thermal conductivity
• G21C 15/00 - Cooling arrangements within the pressure vessel containing the coreSelection of specific coolants

Abstract

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

IPC Classes  ?

• H01M 8/04298 - Processes for controlling fuel cells or fuel cell systems
• G06Q 10/06 - Resources, workflows, human or project managementEnterprise or organisation planningEnterprise or organisation modelling

Abstract

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.

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G06F 30/20 - Design optimisation, verification or simulation
• G06F 17/10 - Complex mathematical operations
• G06N 3/00 - Computing arrangements based on biological models
• G21D 3/00 - Control of nuclear power plant

Abstract

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

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G06N 20/00 - Machine learning
• G06F 17/18 - Complex mathematical operations for evaluating statistical data
• G06F 30/20 - Design optimisation, verification or simulation

Abstract

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.

IPC Classes  ?

• G06Q 10/04 - Forecasting or optimisation specially adapted for administrative or management purposes, e.g. linear programming or "cutting stock problem"
• G06N 3/0442 - Recurrent networks, e.g. Hopfield networks characterised by memory or gating, e.g. long short-term memory [LSTM] or gated recurrent units [GRU]

Abstract

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).

IPC Classes  ?

• G21D 3/00 - Control of nuclear power plant

Abstract

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.

IPC Classes  ?

• G21C 9/02 - Means for effecting very rapid reduction of the reactivity factor under fault conditions, e.g. reactor fuse

Abstract

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.

IPC Classes  ?

• G21D 3/04 - Safety arrangements
• G21C 17/00 - MonitoringTesting
• G21D 1/02 - Arrangements of auxiliary equipment
• G01D 21/02 - Measuring two or more variables by means not covered by a single other subclass

Abstract

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).

IPC Classes  ?

• G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
• G21C 1/03 - Fast fission reactors, i.e. reactors not using a moderator cooled by a coolant not essentially pressurised, e.g. pool-type reactors
• G21C 15/243 - Promoting flow of the coolant for liquids
• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21C 15/14 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts conducting a hot fluidArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts comprising auxiliary apparatus, e.g. pumps, cameras

Abstract

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.

IPC Classes  ?

• G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
• G21C 3/322 - Means to influence the coolant flow through or around the bundles

Abstract

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.

IPC Classes  ?

• G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles

Abstract

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 .

IPC Classes  ?

• G21C 3/356 - Spacer grids being provided with fuel element supporting members
• G21C 3/34 - Spacer grids
• G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles

Abstract

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.

IPC Classes  ?

• G21C 17/032 - Reactor-coolant flow measuring or monitoring

Abstract

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

IPC Classes  ?

• H02K 15/02 - Processes or apparatus specially adapted for manufacturing, assembling, maintaining or repairing of dynamo-electric machines of stator or rotor bodies

Abstract

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.

IPC Classes  ?

• G21C 7/08 - Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of solid control elements, e.g. control rods
• G21C 7/36 - Control circuits
• G21C 7/06 - Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section

Abstract

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.

IPC Classes  ?

• G01N 15/06 - Investigating concentration of particle suspensions
• G01N 15/00 - Investigating characteristics of particlesInvestigating permeability, pore-volume or surface-area of porous materials

Abstract

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.

IPC Classes  ?

• G01N 25/4 -

Abstract

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.

IPC Classes  ?

• G05B 9/03 - Safety arrangements electric with multiple-channel loop, i.e. redundant control systems
• G05D 13/62 - Control of linear speedControl of angular speedControl of acceleration or deceleration, e.g. of a prime mover characterised by the use of electric means, e.g. use of a tachometric dynamo, use of a transducer converting an electric value into a displacement

Abstract

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.

IPC Classes  ?

• G06F 30/27 - Design optimisation, verification or simulation using machine learning, e.g. artificial intelligence, neural networks, support vector machines [SVM] or training a model
• G06F 18/214 - Generating training patternsBootstrap methods, e.g. bagging or boosting
• G06N 3/04 - Architecture, e.g. interconnection topology
• G06N 3/08 - Learning methods

Abstract

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.

IPC Classes  ?

• G21C 3/332 - Supports for spacer grids

Abstract

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.

IPC Classes  ?

• G06F 11/34 - Recording or statistical evaluation of computer activity, e.g. of down time, of input/output operation

Abstract

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.

IPC Classes  ?

• G21C 3/58 - Solid reactor fuel
• G21C 3/06 - CasingsJackets

Abstract

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).

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G06Q 10/04 - Forecasting or optimisation specially adapted for administrative or management purposes, e.g. linear programming or "cutting stock problem"

Abstract

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.

IPC Classes  ?

• G06Q 10/20 - Administration of product repair or maintenance
• G06Q 50/06 - Energy or water supply
• G06N 5/045 - Explanation of inferenceExplainable artificial intelligence [XAI]Interpretable artificial intelligence
• G06N 7/01 - Probabilistic graphical models, e.g. probabilistic networks
• G21D 3/06 - Safety arrangements responsive to faults within the plant

Abstract

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.

IPC Classes  ?

• G21C 3/34 - Spacer grids
• G21C 3/356 - Spacer grids being provided with fuel element supporting members

Abstract

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).

IPC Classes  ?

• B23K 31/02 - Processes relevant to this subclass, specially adapted for particular articles or purposes, but not covered by any single one of main groups relating to soldering or welding
• G21C 7/10 - Construction of control elements

Abstract

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

IPC Classes  ?

• G06Q 10/04 - Forecasting or optimisation specially adapted for administrative or management purposes, e.g. linear programming or "cutting stock problem"

Abstract

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.

IPC Classes  ?

• G01N 27/411 - Cells and probes with solid electrolytes for investigating or analysing of liquid metals
• G21C 17/02 - Devices or arrangements for monitoring coolant or moderator
• G21C 19/07 - Storage racksStorage pools

Abstract

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.

IPC Classes  ?

• G21C 3/08 - CasingsJackets provided with external means to promote heat-transfer, e.g. fins, baffles, corrugations
• G21C 3/22 - Fuel elements with fissile or breeder material in contact with coolant
• G21C 3/07 - CasingsJackets characterised by their material, e.g. alloys

Abstract

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.

IPC Classes  ?

• C04B 35/80 - Fibres, filaments, whiskers, platelets, or the like
• C04B 35/565 - Shaped ceramic products characterised by their compositionCeramic compositionsProcessing powders of inorganic compounds preparatory to the manufacturing of ceramic products based on non-oxides based on carbides based on silicon carbide
• C04B 35/622 - Forming processesProcessing powders of inorganic compounds preparatory to the manufacturing of ceramic products

Abstract

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.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting

Abstract

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).

IPC Classes  ?

• G01N 33/18 - Water
• G06F 17/15 - Correlation function computation

Abstract

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.

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G21C 21/02 - Manufacture of fuel elements or breeder elements contained in non-active casings
• G21C 3/07 - CasingsJackets characterised by their material, e.g. alloys
• G21C 3/10 - End closures
• B23K 9/16 - Arc welding or cutting making use of shielding gas
• B23K 1/00 - Soldering, e.g. brazing, or unsoldering
• B23K 26/12 - Working by laser beam, e.g. welding, cutting or boring in a special environment or atmosphere, e.g. in an enclosure
• B23K 15/10 - Non-vacuum electron beam-welding or cutting
• C04B 37/00 - Joining burned ceramic articles with other burned ceramic articles or other articles by heating

Abstract

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

IPC Classes  ?

• G06T 7/00 - Image analysis

Abstract

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.

IPC Classes  ?

• G06Q 10/04 - Forecasting or optimisation specially adapted for administrative or management purposes, e.g. linear programming or "cutting stock problem"

Abstract

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).

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

• G01P 21/02 - Testing or calibrating of apparatus or devices covered by the other groups of this subclass of speedometers

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

A 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.

IPC Classes  ?

• B23K 37/00 - Auxiliary devices or processes, not specially adapted for a procedure covered by only one of the other main groups of this subclass
• B23K 1/008 - Soldering within a furnace

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

• G21C 7/36 - Control circuits
• G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators

Abstract

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

IPC Classes  ?

• G01R 31/00 - Arrangements for testing electric propertiesArrangements for locating electric faultsArrangements for electrical testing characterised by what is being tested not provided for elsewhere
• F16K 37/00 - Special means in or on valves or other cut-off apparatus for indicating or recording operation thereof, or for enabling an alarm to be given

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

Disclosed 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.

IPC Classes  ?

• G01M 3/40 - Investigating fluid tightness of structures by using electric means, e.g. by observing electric discharges
• G01V 3/11 - Electric or magnetic prospecting or detectingMeasuring magnetic field characteristics of the earth, e.g. declination or deviation operating with magnetic or electric fields produced or modified by objects or geological structures or by detecting devices using induction coils for detecting conductive objects, e.g. firearms, cables or pipes
• G01N 27/90 - Investigating or analysing materials by the use of electric, electrochemical, or magnetic means by investigating magnetic variables for investigating the presence of flaws using eddy currents

Abstract

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

IPC Classes  ?

• G01B 11/00 - Measuring arrangements characterised by the use of optical techniques

Abstract

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

IPC Classes  ?

• G01R 31/34 - Testing dynamo-electric machines
• G01R 19/00 - Arrangements for measuring currents or voltages or for indicating presence or sign thereof
• G01J 5/48 - ThermographyTechniques using wholly visual means

Abstract

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

IPC Classes  ?

• G21D 3/00 - Control of nuclear power plant

Abstract

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

IPC Classes  ?

• B66C 1/12 - Slings comprising chains, wires, ropes, or bandsNets
• B66C 1/14 - Slings with hooks
• B66C 13/08 - Auxiliary devices for controlling movements of suspended loads, or for preventing cable slack for depositing loads in desired attitudes or positions
• B66C 13/16 - Applications of indicating, registering, or weighing devices
• B66C 13/18 - Control systems or devices

Abstract

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).

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

A 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.

IPC Classes  ?

• B01D 33/333 - Filters with filtering elements which move during the filtering operation with individual filtering elements moving along a closed path
• B01D 33/80 - Accessories
• B01D 33/48 - Regenerating the filter material in the filter by flushing, e.g. counter-current air-bumps
• G06F 17/12 - Simultaneous equations

Abstract

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.

IPC Classes  ?

• G01N 1/14 - Suction devices, e.g. pumpsEjector devices
• G21C 17/00 - MonitoringTesting

Abstract

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.

IPC Classes  ?

• G01N 25/20 - Investigating or analysing materials by the use of thermal means by investigating the development of heat, i.e. calorimetry, e.g. by measuring specific heat, by measuring thermal conductivity
• G01N 23/00 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or

Abstract

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.

IPC Classes  ?

• G01N 1/10 - Devices for withdrawing samples in the liquid or fluent state
• G21C 19/28 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core

Abstract

A joint resistance measurement device, comprising: a control unit (160), a constant current source (140), a constant voltage source (180), a first connection end (111), a second connection end (112), a first change-over switch (121), a second change-over switch (122), a first voltage detection unit (131), a second voltage detection unit (132), and a digital-to-analog conversion unit (150). The first connection end (111) and the second connection end (112) are used for being correspondingly connected to an inner core and an outer core of a joint under test; the control unit (160) is configured to output continuous resistance values of said joint when receiving a first trigger instruction. The joint resistance measurement device can effectively improve the test efficiency of the joint and reduce the labor cost.

IPC Classes  ?

• G01R 27/08 - Measuring resistance by measuring both voltage and current

Abstract

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.

IPC Classes  ?

• C04B 37/00 - Joining burned ceramic articles with other burned ceramic articles or other articles by heating
• G21C 3/07 - CasingsJackets characterised by their material, e.g. alloys

Abstract

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.

IPC Classes  ?

• G21C 9/004 - Pressure suppression
• G21C 15/02 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements
• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21C 15/14 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts conducting a hot fluidArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts comprising auxiliary apparatus, e.g. pumps, cameras

Abstract

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.

IPC Classes  ?

• F01D 21/12 - Shutting-down of machines or engines, e.g. in emergencyRegulating, controlling, or safety means not otherwise provided for responsive to temperature
• F01D 21/00 - Shutting-down of machines or engines, e.g. in emergencyRegulating, controlling, or safety means not otherwise provided for
• G21D 3/06 - Safety arrangements responsive to faults within the plant

Abstract

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).

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21D 1/02 - Arrangements of auxiliary equipment

Abstract

The present application relates to the technical field of instrument verification. Provided in the embodiments of the present application is a level gauge calibration device. The level gauge calibration device comprises at least a calibration container (100), a test container (200), a floating block (300), a measurement device (400), an adjustment device (500) and a control device (600). By providing a test container (200) that is in communication with a calibration container (100), the height of the liquid level in the calibration container (100) is the same as the height of the liquid level in the test container (200), so that the height of the liquid level in the calibration container (100) can be obtained simply by measuring the height of the liquid level in the test container (200). By providing the floating block (300) and the measurement device (400), the measurement accuracy can be improved. By providing the adjustment device (500) and the control device (600), the control device (600) can control the adjustment device (500) to act according to the height of the liquid level in the test container (200) determined by the measurement device (400), so as to adjust the height of the liquid level in the calibration container (100). Therefore, through the mutual cooperation of the components, the calibration accuracy of a level gauge is improved.

IPC Classes  ?

• G01F 25/20 - Testing or calibration of apparatus for measuring volume, volume flow or liquid level or for metering by volume of apparatus for measuring liquid level

Abstract

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.

IPC Classes  ?

• G06Q 10/06 - Resources, workflows, human or project managementEnterprise or organisation planningEnterprise or organisation modelling
• G21D 3/16 - Varying reactivity
• G21D 3/00 - Control of nuclear power plant
• G21C 7/08 - Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of solid control elements, e.g. control rods

Abstract

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.

IPC Classes  ?

• G21D 3/04 - Safety arrangements

Abstract

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

IPC Classes  ?

• G21D 3/00 - Control of nuclear power plant

Abstract

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.

IPC Classes  ?

• G05B 13/04 - Adaptive control systems, i.e. systems automatically adjusting themselves to have a performance which is optimum according to some preassigned criterion electric involving the use of models or simulators

Abstract

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).

IPC Classes  ?

• G06F 30/17 - Mechanical parametric or variational design
• G06F 30/23 - Design optimisation, verification or simulation using finite element methods [FEM] or finite difference methods [FDM]
• G06F 111/10 - Numerical modelling
• G06F 119/02 - Reliability analysis or reliability optimisationFailure analysis, e.g. worst case scenario performance, failure mode and effects analysis [FMEA]
• G06F 119/14 - Force analysis or force optimisation, e.g. static or dynamic forces

Abstract

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).

IPC Classes  ?

• G21C 7/10 - Construction of control elements

Abstract

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

IPC Classes  ?

• G21C 17/10 - Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
• G21C 17/104 - Measuring reactivity

Abstract

A thermoelectric power probe, comprising a handheld housing (1), which is provided with a circuit board (5) and an abutting end (11), wherein the abutting end (11) is provided with a detection assembly (12); the detection assembly (12) comprises a first heating unit (21), a second heating unit (22), a first thermocouple (231) and a second thermocouple (241); and junctions (231a, 241a) of the two thermocouples (231, 241) abut against an overlapping region between heat dissipation regions (21f, 22f) of the two heating units (21, 22), and the overlapping region is heated by the two heating units (21, 22) according to a set rule. The thermoelectric power probe uses the two heating units (21, 22), and the difference between the heating temperatures of the two heating units (21, 22) is adjusted, such that the difference between the temperatures of the junctions (231a, 241a) of the two thermocouples (231, 241) in the overlapping region reaches a suitable measurement state, and therefore a measurement result is accurate.

IPC Classes  ?

• G01K 7/02 - Measuring temperature based on the use of electric or magnetic elements directly sensitive to heat using thermoelectric elements, e.g. thermocouples
• G01K 1/08 - Protective devices, e.g. casings

Abstract

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).

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G21C 17/108 - Measuring reactor flux

Abstract

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

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

• G06F 9/445 - Program loading or initiating

Abstract

The present application relates to a 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.

IPC Classes  ?

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

Abstract

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

IPC Classes  ?

• G06F 16/2455 - Query execution

Abstract

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

IPC Classes  ?

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

Abstract

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.

IPC Classes  ?

• G21C 3/62 - Ceramic fuel
• G21C 21/10 - Manufacture of fuel elements or breeder elements contained in non-active casings by extrusion, drawing, or stretching

Abstract

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

IPC Classes  ?

• G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
• G01N 23/00 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or

Abstract

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.

IPC Classes  ?

• G21F 5/08 - Shock-absorbers, e.g. impact buffers for containers
• G21F 5/06 - Details of, or accessories to, the containers

Abstract

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.

IPC Classes  ?

• G06F 30/27 - Design optimisation, verification or simulation using machine learning, e.g. artificial intelligence, neural networks, support vector machines [SVM] or training a model