21 results about "Reactor safety" patented technology

The present application relates to the technical field of reactors, and particularly relates to a calculation method of corrosion product deposition rate of a pressurized water reactor. The calculation method of corrosion product deposition rate of the pressurized water reactor comprises the following steps: obtaining the concentration distribution and Zeta potential distribution of metal ions related to corrosion products; obtaining the electric field force on the particle state corrosion products based on the Zeta potential distribution; tracking the transport trajectory of the particle state corrosion products by using a discrete phase model, wherein the force on the particles in the transport process comprises the electric field force; calculating the particle mass flow rate according to the discrete phase model; and calculating the deposition rate and erosion rate of the particle state corrosion products on the wall surface according to the particle mass flow rate. The present application can be used to solve the problem that the current calculation model cannot accurately predict the CRUD deposition rate, and can improve the authenticity of CRUD growth prediction and the reliability of reactor safety analysis.

This invention discloses a method and apparatus for simulating heat pipe reactor loss-of-heat-sink accidents, including a heat pipe reactor simulation device, a DC power supply, and a secondary loop auxiliary system. The DC power supply is independently located outside the heat pipe reactor simulation device and electrically connected to it. The secondary loop auxiliary system is connected to the piping of the heat pipe reactor simulation device. This solution, based on "scaled modeling + precise modeling + safe substitution," constructs a heat pipe reactor accident simulation system that combines realism, safety, and economy. It avoids nuclear risks while accurately reproducing key scenarios of coolant loss accidents, solving the pain points of real reactor testing. It provides an efficient technical means for the safe design, optimization, and accident response of heat pipe reactors, and is of great significance for improving the safe operation level of nuclear reactors.

A method for analyzing the bubbling behavior of high-fuel-consumption fuel under a transient high-temperature condition relates to the technical field of reactor safety analysis, and comprises the following steps: establishing a three-dimensional model; subdividing the three-dimensional model to obtain a computational domain, and performing physical property assignment on each unit grid of the computational domain; an irradiation test process is simulated, and the fission rate of fuel of each unit grid is calculated; within the irradiation time, the fuel consumption is calculated and updated, and irradiation swelling and creep behaviors of the fuel core are calculated; calculating the total amount of fission gas released by the fuel core body; adding a virtual bubbling core in the three-dimensional model; based on the volume expansion increment of the fuel core body and the true stress of the cladding, simulating and calculating the swelling behavior of the fuel core body, the creep behavior of the cladding and the plastic deformation process of the cladding in the gradual heating process of the fuel by setting external heating boundary conditions; the method is used for solving the problem that a traditional method cannot accurately analyze bubbling changes in a fuel assembly.

The application discloses an advanced safety injection system and a pressurized water reactor, and belongs to the nuclear power field. The advanced safety injection system comprises an integrated pressure water tank, the integrated pressure water tank comprises a CMT cavity and an ACC cavity which are separated by a conical wall extending vertically along an axis, the ACC cavity is arranged outside the CMT cavity, the cross-sectional area of the ACC cavity gradually decreases from top to bottom, and upper and lower through holes are arranged on the conical wall to communicate the CMT cavity and the ACC cavity. The ACC cavity is provided with pressure gas at the top; the CMT cavity is provided with a top inlet communicated with a reactor primary cooling system at the top end and is provided with an injection pipeline for injecting coolant into a reactor core at the bottom end. The advanced safety injection system has simple structure, can play the CMT and ACC functions under different accident conditions by using a single water tank, effectively reduces the manufacturing and maintenance costs, and improves the safety of the reactor.

This utility model discloses an inert gas reactor safety protection device, comprising a transfer chamber connected to a nitrogen supply pipeline and a nitrogen purging pipeline. The bottom of the transfer chamber is provided with a connecting pipeline to the reactor. An emergency nitrogen supply pipeline is also provided on the transfer chamber. Several gas transfer tanks are located inside the transfer chamber, and each gas transfer tank is connected to an external exhaust pipeline equipped with a flame arrestor valve. The transfer chamber is equipped with a central control panel. This utility model utilizes three parts—inert gas protection, purging, and emergency supply—and employs multiple functional valves, along with a highly integrated central control panel and online oxygen content detection equipment. It automatically supplies nitrogen for purging based on the internal oxygen content, features multi-module safety self-locking control, and includes an emergency nitrogen supply function during the oxidation process, improving the safety factor and achieving intrinsic safety. The entire system is easy to observe and operate, with a high degree of automation.

A method and system for constructing a reactor system analysis model based on a graphical module includes the following steps: Step 1: Divide the proposed reactor system prototype object into thermal-hydraulic components, control system components, and reactor neutron physics components; Step 2: Define and draw the graphical primitives of the components divided in Step 1; Step 3: Adjust the component primitives and the link relationships between them; Step 4: Edit the attributes of the component primitives; Step 5: Visualize and store the graphical structure and data of the component primitives, forming multiple reactor system safety analysis models; Step 6: Encapsulate and save the formed reactor system safety analysis models. This application reduces the professional requirements and difficulty of constructing reactor system safety analysis models, improves the work efficiency of modelers, and shortens the reactor safety analysis cycle.

This invention discloses a method for preparing a wear-resistant thermal barrier coating slurry based on Gemini surfactant in colloidal form. The method involves synthesizing an asymmetric Gemini surfactant, using Al(NO₃)₃ and zirconium oxynitrate as ceramic matrix raw materials, and adding additives such as benzyl dimethylamine. The target slurry is prepared through microemulsion preparation, mixing and precipitation, filtration and washing. This invention solves the technical problems of existing new energy vehicle battery box coatings, which struggle to simultaneously achieve wear resistance and thermal barrier performance, and have insufficient resistance to various media. The resulting slurry forms a coating with a wear amount as low as 0.142g, can withstand kerosene and gasoline corrosion at 252℃ for 14 days without abnormalities, and has a room temperature drying time of 24-60 hours. It significantly improves the corrosion inhibition and safety of battery boxes, and is suitable for high-temperature wear protection in various scenarios such as wind turbine blades, nuclear reactor containment vessels, dam structures, and turbines.

A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs, each of the functionally independent modules comprising a digital module or a combination digital and analog module, an analog module electrically coupled to one or more of the functionally independent modules, and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

The present disclosure provides an isotope irradiation safety analysis method and device, electronic equipment and storage medium, relating to the technical field of nuclear power generation, by comprehensively considering the irradiation requirements of target isotopes, side reflection layer hole configuration parameters, and material properties and structural parameters of irradiation targets, two types of calculation models corresponding to unloaded and loaded irradiation target states are constructed to obtain baseline and perturbed core power distribution, and then the degree of core power distortion is quantitatively characterized, and based on the comparison results, the irradiation target design parameters and / or hole configuration parameters are adjusted to iteratively optimize the irradiation scheme, so that the problem that the prior art lacks a systematic analysis method of comprehensive key factors, can only qualitatively describe or simply estimate the core power distortion, and cannot meet the reactor safety evaluation requirements, can be solved, and the technical effects of accurately quantifying the degree of core power distortion and ensuring that the isotope irradiation scheme meets the safety requirements of the reactor are achieved.

Disclosed is a method of testing a leak rate of a reactor containment building, the method including: a division step of dividing the reactor containment building into an upper containment area and a lower containment area by a spectacle flange; a measurement step of measuring the leak rate of the containment building after pressurizing the inside of the containment building at a maximum pressure; and an identification step of identifying whether the leak rate measured in the measurement step satisfies an allowable standard for the leak rate of the containment building.

The application discloses a high-temperature gas cooled reactor safety analysis method and device, electronic equipment and storage medium, relates to the nuclear reactor technical field, and includes: constructing a neural network model; pre-training and fine-tuning the model based on physical simulation data and experimental data; generating virtual training samples for extreme accident conditions to enhance the model generalization ability; inputting the real-time monitored irradiation loop operating parameters into the trained model, and outputting the target channel temperature prediction value and safety margin evaluation result. The problems of high calculation cost caused by dependence on complex physical models, difficulty in real-time prediction, insufficient experimental data leading to weak model generalization ability, and inability to quickly provide emergency decision support under extreme conditions are solved. The technical effects of accurately predicting temperature response, improving the adaptability of the model to different conditions, and providing efficient and real-time data support for high-temperature gas cooled reactor irradiation loop safety analysis and emergency decision-making are achieved.

To efficiently dismantle bottomed cylindrical members. [Solution] A method for cutting a bottomed cylindrical member provided inside a reactor containment vessel, having a body portion extending vertically and a lower closing portion that closes the lower end of the body portion, comprising the steps of driving a disc saw to rotate the cutting disc of the disc saw, and feeding the disc saw along the surface of the body portion with the cutting disc penetrating the body portion to cut the body portion, wherein in the step of rotating the cutting disc, in the step of cutting the body portion, the disc saw is driven in a rotational direction in which the outer circumference of the cutting disc in contact with the body portion transitions in the following order: radially outward of the bottomed cylindrical member, forward in the direction of feeding the disc saw, radially inward of the bottomed cylindrical member, and rearward in the direction of feeding the disc saw.

An example for continuous removal of fission products from a molten-salt fueled nuclear reactor enclosed in a reactor containment may include separating actinides from other fission products flowing out of the reactor, returning the separated actinides back to the reactor to be consumed, and removing the other fission products out of the containment while the reactor is operating. In various embodiments, the reactor can be a critical reactor, a subcritical (e.g., accelerator-drive) reactor, or another type of reactor.

The invention belongs to the technical field of nuclear reactor design, and particularly relates to a passive starting method based on in-reactor detection. The method comprises the following steps: arranging a neutron detector in a reactor core fuel assembly, and obtaining an effective counting rate signal not lower than 0.5 cps by the neutron detector when the reactor is in an initial starting state, a coolant is at normal temperature, all control rods are inserted into a reactor core and the boron concentration is shutdown boron concentration; in the subsequent starting process, along with the rising of the coolant temperature, the reduction of the control rod lifting reactor core or the boron concentration, the continuous increase of the counting rate of the neutron detector, no detection blind area exists, and the introduction of unexpected reactivity is monitored in real time, so that the reactor safely and effectively reaches the critical state; and meanwhile, in the starting process, the rod lifting and boron dilution reactivity introduction process is guided by combining the counting rate change condition of the neutron detector. According to the invention, the neutron detector arranged in the reactor core fuel assembly is utilized to realize non-blind area monitoring in the whole starting process, so that safe and efficient passive starting of the reactor is ensured.

A numerical simulation analysis method for the swelling and bursting of pressurized water reactor fuel rod cladding, relating to the field of reactor safety analysis technology, includes: obtaining constant parameters of the fuel rod cladding; establishing and meshing a geometric model; assigning physical property values ​​to each element mesh; calculating the temperature distribution of the fuel rod cladding and determining whether a phase transition has occurred; if a phase transition has occurred, calculating the phase transition process of the cladding material; if the phase transition is complete or has not occurred, calculating the creep and plastic behavior of the cladding material; updating the effective stress and effective strain rate of the cladding material, and then calculating the yield stress of each element mesh; determining whether both the effective strain rate and yield stress have reached the corresponding failure thresholds; if neither has reached the threshold, repeating the calculation; if either reaches the failure threshold, performing a dynamic bursting process calculation of the cladding until the pressure inside and outside the cladding is balanced, and simulating the dynamic bursting process of the cladding; this method addresses the limitations and inaccurate results of traditional bursting analysis.

The present application relates to the technical field of nuclear power plant nuclear reactor safety, and is a space heat pipe nuclear reactor design method and computer program product based on shutdown safety physical modeling, which solves the problem that the existing space heat pipe nuclear reactor design method is insufficient in consideration and cannot take into account the safety, power capacity and lightweight requirements; the present application comprises the following steps: establishing a mass model, a safety model and a power model of the nuclear reactor; inputting the constraint conditions of the safety model into the model to make the design meet the shutdown safety requirements; inputting the constraint conditions of the mass model and the power model into the corresponding models respectively and performing framework optimization; the present application constructs an interpretable shutdown neutronics equivalent model, takes the influence of the absorber drum assembly and the reflector on the shutdown criticality as a hard safety constraint, and on this basis, jointly designs the reactor power level and the system mass.

The present invention belongs to the technical field of nuclear reactor design, and particularly relates to an in-core detection-based passive start-up method. A neutron detector is provided inside a reactor core fuel assembly. In an initial state of reactor start-up, a coolant is at room temperature, all control rods are inserted into a reactor core and a boron concentration is a shutdown boron concentration, the neutron detector obtains an effective count rate signal of not less than 0.5cps. During subsequent start-up, as the temperature of the coolant increases, the control rods are withdrawn from the reactor core or the boron concentration decreases, the neutron detector count rate exhibits a continuous increase, no detection blind zone is present, and real-time monitoring of inadvertent reactivity insertion is achieved, thereby enabling a reactor to reach a critical state safely and effectively. Additionally, by taking into account variations of the neutron detector count rate during the start-up, reactivity insertion processes of rod withdrawal and boron dilution are guided. The present invention uses the neutron detector arranged in the reactor core fuel assembly to realize blind-zone-free monitoring during the entire start-up process, thereby ensuring safe and efficient passive stat-up of the reactor.

The application discloses a prediction method and device of effective multiplication factor uncertainty, and a reactor safety evaluation method and system, and relates to the technical field of nuclear reactors. The prediction method comprises the following steps: obtaining a single-parameter sampling data set of a target reactor; determining c groups of effective multiplication factor prediction values corresponding to the target reactor according to a target effective multiplication factor prediction model and the single-parameter sampling data set; and determining c effective multiplication factor uncertainty prediction values corresponding to the target reactor according to the c groups of effective multiplication factor prediction values. According to the embodiment of the application, the prediction speed and the prediction accuracy of the effective multiplication factor uncertainty can be improved.

The invention belongs to the technical field of nuclear reactor design, and particularly relates to a passive starting method for arranging a source range detector in a reactor reflecting layer. A source range detector is arranged in a reactor reflecting layer, and the source range detector arranged in the reactor reflecting layer monitors introduction of unexpected reactivity in real time in the whole process from reactor charging completion to critical starting, so that a reactor safely and effectively reaches a critical state. When a reactor is started in an initial state, a coolant is at normal temperature, all the control rods are inserted into a reactor core and the boron concentration is shutdown boron concentration, the detector obtains an effective counting rate signal not lower than 0.5 cps. According to the invention, by using the source range detector arranged in the reflecting layer of the reactor, non-blind area monitoring is realized in the whole starting process, and safe and efficient passive starting of the reactor is ensured. According to the invention, the dependence of the reactor starting on the external starting neutron source is avoided, the construction cost is reduced, and the generation of radioactive solid wastes related to the neutron source is eliminated.

The embodiments of this application relate to the field of nuclear reactor monitoring, and particularly to a method for determining the power of heat generated by the decay of spent fuel assemblies in a sodium-cooled fast reactor. The method includes the following steps: S10: placing the spent fuel assembly of the sodium-cooled fast reactor whose decay heat is to be determined in a measurement space; S20: placing a coolant in the measurement space, the coolant being configured to conduct away a portion of the decay heat generated by the spent fuel assembly; S30: determining the average temperature of the walls of the measurement space, determining the power of heat dissipation from the measurement space and the power of heat conducted away by the coolant; S40: determining the power of heat generated by the decay of the spent fuel assembly based on the power of heat dissipation from the measurement space and the power of heat conducted away by the coolant. This makes the decay heat power of the spent fuel assembly more accurate, reduces the uncertainty level, thereby reducing the design conservatism margin of the decay heat extraction system, and avoids safety problems such as coolant overcooling, improving reactor safety and economy.

The invention belongs to the technical field of nuclear engineering equipment health management, and particularly relates to a high-frequency original data processing method based on a reactor system acquisition device. The data acquisition layer is composed of FTP drive acquisition, OPC UA drive acquisition and Modbus drive acquisition and is used for realizing acquisition of high-frequency original data of a reactor system; the data processing layer is used for carrying out preprocessing, feature value extraction and feature value application on the collected high-frequency original data; the data distribution layer is used for processing data transmission of different priorities through data transmission channel redundancy and differentiation according to the data transmission priority requirements of a reactor system so as to realize hierarchical transmission of the data and optimize transmission performance; and the data distribution layer process comprises formulating a data priority standard and configuring a data sending channel. The sensor data transmission efficiency of the acquisition device of the reactor system is improved, the reliability and the real-time performance of the data are ensured, and reliable guarantee is provided for safe operation of the reactor.