18 results about "Nuclear reactor coolant" patented technology

The invention belongs to the technical field of optical fiber high-temperature sensing, and discloses a high-temperature-resistant and high-pressure-resistant optical fiber cascade refractive index-temperature double-parameter integrated sensor which comprises a packaging structure, an optical fiber sensing device and a transmission optical fiber. The optical fiber sensing device comprises a photonic crystal fiber, a hollow Bragg fiber, a connecting fiber and a hollow capillary fiber which are connected in sequence; the hollow capillary optical fiber at the tail end of the optical fiber sensing device is connected with the transmission optical fiber; the optical fiber sensing device is arranged in the packaging structure; an outer air hole is formed in a position, opposite to the photonic crystal fiber, of the side wall of the packaging structure; and the optical fiber sensing device transmits a sensing signal to the outside of the packaging structure through the transmission optical fiber. According to the invention, refractive index-temperature in-situ double-parameter synchronous high-precision measurement capable of resisting high temperature of 1100 DEG C can be realized, and the method has important application value in the fields of aero-engine combustion chamber combustion state monitoring, nuclear reactor coolant concentration analysis, oil exploitation component analysis and the like.

The invention belongs to the technical field of nuclear power, and discloses a degassing system and a degassing method for a nuclear reactor coolant. The degassing system of the nuclear reactor coolant comprises a hollow fiber membrane degassing device, a coolant storage device connected with a coolant input end of the hollow fiber membrane degassing device, and a vacuum device connected with a nitrogen output end of the hollow fiber membrane degassing device, the hollow fiber membrane degassing device comprises at least one vertically arranged degasser; the degasser comprises a coolant channel and a membrane wire channel which are arranged in the vertical direction; coolant flows in the coolant channel from bottom to top, and nitrogen flows in the membrane filament channel from top to bottom. The hydrophobic characteristic of the hollow fiber membrane is utilized, and the membrane degassing technology is combined with vacuum suction and nitrogen purging to achieve the deep degassing target. According to the degassing system for the nuclear reactor coolant, the coolant flows from bottom to top in cooperation with nitrogen flowing from top to bottom, so that the whole system operates reliably for a long time, and energy consumption is optimized.

The present application belongs to the field of nuclear power plant reactor coolant pump motor unit installation and test, and particularly relates to a nuclear reactor coolant pump motor rotor reset measuring device and method, which comprises: a detection assembly and a data processing assembly; the detection assembly is multiple sets, and is arranged on the upper slip rotor and the lower flywheel of the motor, respectively; the data processing assembly comprises a control box and a remote control PC end; the detection assembly, the control box and the remote control PC end are connected in sequence, the detection assembly measures and collects the rotor position, generates measurement and collection data, and transmits the data to the control box; the control box receives the measurement and collection data transmitted by the detection assembly, processes the data, generates measurement result data, and transmits the data to the remote control PC end; and the remote control PC end receives the measurement result data transmitted by the control box, and displays the data. The device has simple structure and convenient operation, and can realize the rotor reset after the no-load test of the reactor coolant pump motor.

The invention belongs to the technical field of nuclear reactor coolant purification and metallurgy, and particularly relates to an active liquid lead-based metal impurity trapping device. The embodiment of the invention provides an active liquid lead-based metal impurity trapping device which comprises a narrow spoke type crawler belt, an impurity collecting tank, a driving motor and a driving mechanism, and the driving mechanism drives the narrow spoke type crawler belt to move, so that impurities are brought out of a free liquid level and enter the impurity collecting tank to be trapped. The trapping device is compact in structure, reliable in operation, capable of continuously working on line and particularly suitable for the field of liquid lead-based metal cooling tests, liquid metal can effectively pass through gaps by adopting a narrow spoke type crawler belt, a good collecting effect is achieved on granular, flaky and bulk impurities at a free liquid level, non-stop online continuous operation is achieved, and the working efficiency is improved. The purification efficiency is greatly improved, and long-term stable operation of a main system is guaranteed.

The invention belongs to the technical field of material testing and corrosion and abrasion evaluation, and particularly relates to a corrosion and abrasion testing device for a lead-bismuth main pump impeller. The device comprises a coupler, a supporting frame, a crucible, a base, a motor, a bearing, a sealing device, a sealing cavity, a shafting part, a part hanging piece and an impeller shaft frame. According to the method, the service performance of different materials under the real operation condition of the lead-bismuth main pump impeller can be simulated, and a test basis is provided for selection of the lead-bismuth main pump impeller material. The device is used for simulating the corrosion and abrasion conditions of an impeller in a nuclear reactor coolant system in a high-temperature lead-bismuth liquid environment, so that the corrosion resistance and abrasion resistance of different materials in the lead-bismuth environment are evaluated, and data support is provided for selection and improvement of materials of the impeller for a lead-bismuth pump.

The present disclosure relates to an apparatus for detecting nuclear reactor coolant leaks that is capable of maintaining radiation detection reliability and the integrity of a radiation detector from high-dose background radiation inside a nuclear reactor when a coolant leaks from the primary coolant system of the nuclear reactor. The apparatus of the present disclosure includes a shield having an internal space into which sample air containing a coolant leaking from a primary coolant system pipe of a nuclear reactor is introduced. The shield is configured to close the exposed internal space or expand the internal space through detachable coupling. The apparatus of the present disclosure further includes a first sensor disposed in the internal space to obtain a radiation measurement signal from the sample air. The apparatus of the present disclosure further includes a discriminator for determining whether a coolant is leaking based on the radiation measurement signal.

A coast-down apparatus for a reactor coolant pump (RCP) of a small modular nuclear reactor of the present invention comprises: a solenoid valve; a rotor including an impulse blade and installed on a rotary shaft; and an energy storage unit for rotation-driving the impulse blade, wherein: the rotary shaft of a nuclear reactor coolant pump (RCP) is provided with a rotation detection unit for detecting whether to rotate; the rotor installed on the rotary shaft is provided with a motor, an impeller, and the impulse blade for receiving energy from the energy storage unit to rotate; when power of the nuclear reactor coolant pump (RCP) is lost, the solenoid valve performs a fail-open operation; a power supply unit supplies, when power is not supplied to the reactor coolant pump (RCP), power to each of the motor and the solenoid valve through a power supply line, and when the solenoid valve is opened, receives energy for rotating the impulse blade of the reactor coolant pump (RCP) from the energy storage unit through an energy supply pipe and a nozzle.

The invention discloses a laser cladding coating with an in-situ synthesis TiC and Y-Ti-O composite reinforcement phase and a preparation method of the laser cladding coating, and belongs to the technical field of alloy protective coatings. The coating is cladded on the surface of the steel substrate; the coating comprises a Fe2AlCr alloy matrix phase and an in-situ synthesis reinforced phase which is dispersed and distributed in the Fe2AlCr alloy matrix phase, the in-situ synthesis reinforced phase comprises a TiC phase and a Y-Ti-O composite phase existing in the TiC phase. According to the invention, TiC is obviously refined and spheroidized by a unique composite structure containing phases in phases, and the distribution of TiC is improved, so that the hardness, the wear resistance, the bonding strength with a matrix and the high-temperature liquid lead and bismuth corrosion resistance of the coating are synergistically improved, and the coating is particularly suitable for structural material protection of a fourth-generation nuclear reactor coolant system.

The invention belongs to the field of computational mechanics, and provides a near-field dynamics thermal-mechanical coupling method for analyzing structural fracture of a pressure vessel, which adopts a forward Euler algorithm and a dynamic relaxation algorithm for staggered solution to avoid inconvenience of stiffness matrix change and the like caused by degree-of-freedom change of an implicit method. Meanwhile, according to the method, the axisymmetric near-field dynamic model and the elastoplasticity-creep coupling constitutive integral are combined, and the capacity of the near-field dynamic model in the aspect of analyzing and predicting the elastoplasticity-creep coupling fracture of the axisymmetric structure after ablation is effectively improved. According to the axial symmetry fracture model coupled with the elastoplasticity-creep effect, the fracture behavior of the structure is effectively represented, and the difficulty of fracture analysis under the elastoplasticity-creep coupling effect is overcome. As a new numerical solution format, the method provided by the invention can be realized through simple programming, the numerical implementation complexity is reduced, and favorable help is provided for the research under the serious loss accident of the nuclear reactor coolant.

The embodiment of the invention discloses a performance test method and system for a nuclear reactor coolant pump, and relates to the technical field of nuclear reactors, and the method comprises the steps: obtaining a first measurement value sequence of a main pump in a nuclear reactor in a power-off state, and a second measurement value of the main pump in a power-off instant state, the first measurement value sequence comprises a plurality of first measurement values; determining an initial descent curve of the coolant flow in the loop according to the first measured value sequence and the second measured value; determining a target descent curve according to the initial descent curve and the coolant steady-state flow of each coolant loop in the loop of the nuclear reactor working under different powers; according to the target descent curve, the first preset running-down flow curve and the second preset running-down flow curve, whether the performance test result of the main pump meets the safety analysis requirement or not is determined, and the accuracy of the running-down performance test of the main pump can be improved.

The invention relates to a method for judging an inert flow test of a coolant pump of a nuclear reactor. The method comprises the following steps of: 1, collecting flows of three loops and rotating speeds of the coolant pump of the reactor of the three loops; 2, stopping all reactor coolant pumps at the same time; 3, recording the flows and the pump rotating speeds of the three loops at different moments by taking simultaneous shutdown of the three reactor coolant pumps as a zero moment; 4, calculating the reactor coolant flow test report recording condition according to heat balance, wherein the three loop flows are Q1, Q2 and Q3 respectively when the last fuel circulation reactor coolant system operates stably; 5, preliminarily judging whether the flow of each loop meets the requirement or not; 6, preliminarily judging whether the total flow of the coolant in the core of the primary-loop reactor meets the requirement or not; seventhly, the rotation speed drop of each loop pump is compared with a design curve; and 8, finally determining a test result. The method is used for judging whether the inert flow test of the second-generation pressurized water reactor coolant pump or the second-generation pressurized water reactor coolant pump is qualified or not.

The invention relates to a dynamic self-calibration method, device and equipment for a silk screen sensor, a medium and a program product. The method comprises the steps that current signals of all nodes of a silk screen sensor are collected in real time; according to the current signal of each node, carrying out frequency distribution statistics on the current value sequence of each node to obtain a histogram containing current distribution peak pattern characteristics; extracting a peak value of current distribution in the histogram as a liquid phase background characteristic value, and calculating normalized conductivity based on the liquid phase background characteristic value; and determining void fraction space-time distribution data based on a Maxwell equation and the normalized conductivity. Therefore, the signal characteristics of the gas-liquid two-phase flow can be directly utilized, the liquid-phase background characteristic value is extracted through frequency distribution statistics, the normalized conductivity is calculated, and a calibration relation is established. The method does not need to interrupt the experiment for calibration, can adapt to the temperature change, remarkably improves the measurement efficiency and precision, and is suitable for two-phase flow measurement in severe environments such as nuclear reactor coolant monitoring.

The present application belongs to the technical field of nuclear reactor fuel element thermal hydraulic, and particularly relates to a method for realizing non-uniform temperature distribution of a fuel element base body. The method comprises: arranging thermocouples for temperature measurement at different positions on the side surface of the base body and the surfaces of the electric heating element and the metal gas cooling pipe, and monitoring the temperature distribution of the fuel element base body structure in real time. The method provided by the present application considers the influence of the coolant flow in the real nuclear reactor fuel element on the temperature distribution of the base body, divides the different channels in the fuel element base body into heating channels and cooling channels by function division, integrates the electric heating element in the heating channels to simulate the heat release of the nuclear reactor fuel element, integrates the metal gas cooling pipe in the cooling channels to simulate the cooling effect of the nuclear reactor coolant, and thus realizes more realistic simulation of the temperature field distribution of the nuclear reactor.

Combined cleanup and heat sink systems work with nuclear reactor coolant loops. Combined systems may join hotter and colder sections of the coolant loops in parallel with any steam generator or other extractor and provide optional heat removal between the same. Combined systems also remove impurities or debris from a fluid coolant without significant heat loss from the coolant. A cooler in the combined system may increase in capacity or be augmented in number to move between purifying cooling and major heat removal from the coolant, potentially as an emergency cooler. The cooler may be joined to the hotter and colder sections through valved flow paths depending on desired functionality. Sections of the coolant loops may be fully above the cooler, which may be above the reactor, to drive flow by gravity and enhance isolation of sections of the coolant loop.

This invention discloses a quantitative prediction method for scouring corrosion of flow components in lead-bismuth main pumps, belonging to the field of corrosion calculation technology for nuclear reactor coolant circulation equipment. The invention employs a Kays94 Prandtl number model for turbulent flow, modified for computational fluid dynamics simulation, to obtain the surface temperature, oxygen concentration, flow velocity, and wall shear stress of the flow components. Then, a coupled scouring corrosion model is constructed, incorporating oxide layer growth, reduction removal, and flow erosion removal. Finally, the flow field parameters are input into the coupled model and solved using the oxygen transport equation to quantitatively calculate the magnetite layer thickness, spinel layer thickness, and steel matrix loss. This invention achieves a leap from qualitative judgment to quantitative prediction, directly outputting parameters such as corrosion rate and thickness distribution, providing a quantitative basis for lead-bismuth pump structure optimization, life assessment, and corrosion control.

The embodiment of the application discloses a kind of performance test method and system of nuclear reactor coolant pump, it is related to nuclear reactor technical field, method includes: obtaining the first measurement sequence of main pump in nuclear reactor under the condition of power failure, and the second measurement value of main pump under the condition of power failure moment, first measurement sequence includes multiple first measurement values;According to first measurement sequence and second measurement value, determine the initial drop curve of coolant flow in primary loop;According to initial drop curve, and the coolant steady flow of each coolant loop in primary loop under different power of nuclear reactor, determine target drop curve;According to target drop curve, first preset coasting flow curve and second preset coasting flow curve, determine whether the performance test result of main pump meets the safety analysis requirement, can improve the accuracy of main pump coasting performance test.

One or more embodiments of the invention provide a device and a method for repeatedly unloading and loading a coolant of a small and miniature nuclear reactor. The device comprises a reaction kettle for bearing the small nuclear reactor, a storage tank and a steam generator, the storage tank is internally provided with a liquid alloy coolant, the liquid alloy coolant is supplemented from a cooling hollow rod which is arranged above the storage tank and penetrates through the interior of the storage tank, and a corresponding measuring facility and a safety protection device are further arranged above the storage tank. The pressure control port controls the internal pressure by introducing argon and discharging argon, and a liquid alloy coolant discharging port is formed below the storage tank; the safety guarantee is ensured when the liquid alloy coolant is unloaded and loaded, the storage tank, the phase change, the cooling and the reaction kettle cover the liquid coolant, the coolant is added or unloaded, the washing of the high-temperature, high-pressure and high-flow-speed coolant is avoided, and the chemical action of fission products and the complex mechanical load are borne at the same time.

A guide vane body with a vortex suppression structure ring and an axial-flow nuclear reactor coolant pump (RCP), including a rim cover plate 4, a hub cover plate 5 arranged coaxially within the rim co