CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
Inventor
Xue, Jiaxiang
Wu, Lixiang
Liao, Yehong
Yang, Rongkun
Xiao, Ling
Abstract
Disclosed in the present invention are a silicon carbide composite cladding and a preparation method therefor. The preparation method comprises: S1, providing a silicon carbide slurry; S2, enabling continuous silicon carbide fibers to pass through the silicon carbide slurry, so that the continuous silicon carbide fibers are impregnated with the silicon carbide slurry; S3, braiding the continuous silicon carbide fibers on the surface of an inner pipe to form a silicon carbide fiber braided layer; and S4, performing densification treatment on the silicon carbide fiber braided layer and an interface layer between the silicon carbide fiber braided layer and the inner pipe by means of chemical vapor infiltration. According to the preparation method for the silicon carbide composite cladding, the silicon carbide fiber braided layer impregnated with the silicon carbide slurry is composited on the outer surface of the inner pipe, and then densification treatment is carried out, to reduce pore defects in and between fiber bundles, improve the density, regulate the interface between the fiber and the inner pipe, and improve the toughness of the silicon carbide fiber braided layer, thereby obtaining a silicon carbide composite cladding having high strength, high thermal conductivity and high density.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Lu, Zhiwei
Xue, Jiaxiang
Wu, Lixiang
Liu, Yang
Zhang, Xiansheng
Chen, Yinghong
Yang, Rongkun
Liao, Yehong
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.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Xue, Jiaxiang
Wu, Lixiang
Liao, Yehong
Ren, Qisen
Yang, Rongkun
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.
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
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Xue, Jiaxiang
Wu, Lixiang
Liu, Yang
Liao, Yehong
Ren, Qisen
Zhai, Jianhan
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.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Wu, Lixiang
Liu, Yang
Xue, Jiaxiang
Zhai, Jianhan
Liao, Yehong
Ren, Qisen
Zhang, Xiansheng
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.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Zhang, Feifei
Chen, Yaxuan
Cui, Dawei
Duan, Chengjie
Lin, Jiming
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.
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
7.
SILICON CARBIDE CLADDING CONNECTION MATERIAL, SILICON CARBIDE CERAMIC CONNECTOR AND MANUFACTURING METHOD THEREFOR, AND DEVICE HOUSING
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Wu, Lixiang
Xue, Jiaxiang
Liao, Yehong
Ren, Qisen
Zhai, Jianhan
Liu, Yang
Ma, Haibin
Zhang, Xiansheng
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.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Ding, Peng
Li, Wenhuai
Chen, Shu
Xia, Wenqing
Yu, Fengwan
Duan, Chengjie
Cui, Dawei
Lin, Jiming
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.
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
9.
NUCLEAR REACTOR DESIGN SCHEME OPTIMIZATION METHOD AND APPARATUS, COMPUTER DEVICE, AND MEDIUM
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Ding, Peng
Chen, Shu
Li, Wenhuai
Xia, Wenqing
Yu, Fengwan
Duan, Chengjie
Cui, Dawei
Lin, Jiming
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.
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
10.
ON-LINE DETECTION DEVICE AND METHOD FOR IMPURITIES IN LEAD-BISMUTH COOLANT
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
Inventor
Zeng, Xian
Hu, Chen
Sun, Zhaoxuan
Zou, Qing
Zhang, Yong
Zhao, Yuan
Luo, Yiwei
Duan, Chengjie
Cui, Dawei
Lin, Jiming
Abstract
An on-line detection device (1) and method for impurities in a lead-bismuth coolant (3). The on-line detection device (1) for impurities in the lead-bismuth coolant (3) comprises a measurement channel (10), a pulse laser (20), an optical lens (30), and a spectrometer (40); the measurement channel (10) is connected between the top cover of a pressure vessel (2) and the lead-bismuth coolant (3) inside the pressure vessel (2); the pulse laser (20) and the spectrometer (40) are both provided above the top cover of the pressure container (2); the optical lens (30) is arranged in the measurement channel (10) and is located on an emitting light path of the pulse laser (20), and focuses a pulse laser beam emitted by the pulse laser (20) and refracts same to the surface of the lead-bismuth coolant (3) to generate plasma; the spectrometer (40) is connected to the measurement channel (10), and acquires the plasma to obtain the spectral line intensities of impurity elements. By means of cooperation of the measurement channel (10), the pulse laser (20), and the spectrometer (40), the content of impurities in the lead-bismuth coolant (3) can be accurately detected by LIBS, the urgent need for multi-element, in-situ, and remote online measurement and analysis of impurities in a lead-based alloy is met, and the online monitoring requirement for reactor operation is met.
G01N 21/63 - Systems in which the material investigated is excited whereby it emits light or causes a change in wavelength of the incident light optically excited
G21C 17/025 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators for monitoring liquid metal coolants
11.
MEMBRANE FILTERING MATERIAL, PREPARATION METHOD THEREFOR, AND USE THEREOF IN TREATMENT OF AEROSOL
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Zhao, Yuan
Zeng, Xian
Zou, Qing
Hu, Chen
Luo, Yiwei
Duan, Chengjie
Lin, Jiming
Abstract
2322 hybrid nanofiber membrane. The present invention has a low preparation cost and simple operation; the prepared inorganic nanofiber membrane can be used to efficiently filter an aerosol under high-temperature conditions; and particularly, the filtering efficiency for a polonium aerosol can reach 99%.
B01D 46/54 - Particle separators, e.g. dust precipitators, using ultra-fine filter sheets or diaphragms
D04H 1/728 - Non-woven fabrics formed wholly or mainly of staple fibres or like relatively short fibres characterised by the method of forming fleeces or layers, e.g. reorientation of fibres the fibres being randomly arranged by electro-spinning
D01D 5/00 - Formation of filaments, threads, or the like
D01F 9/08 - Man-made filaments or the like of other substancesManufacture thereofApparatus specially adapted for the manufacture of carbon filaments of inorganic material
D01F 8/18 - Conjugated, i.e. bi- or multicomponent, man-made filaments or the likeManufacture thereof from other substances
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Wu, Lixiang
Xue, Jiaxiang
Liao, Yehong
Ren, Qisen
Ma, Haibin
Zhai, Jianhan
Zhang, Yongdong
Gao, Siyu
Abstract
An induction heating-based silicon carbide cladding connecting method and a silicon carbide cladding. The induction heating-based silicon carbide cladding connecting method comprises: adding a connecting material and a dispersing agent into an organic solvent for dispersing to form a mixed slurry; evenly spreading the mixed slurry on connecting surfaces of an SiC end plug and/or an SiC cladding tube, and aligning the connecting surfaces to match the SiC end plug and the SiC cladding tube to form a connecting structure; curing the connecting structure under a protective atmosphere, the mixed slurry being cured to form a connecting layer; placing the connecting structure on an induction heating device for induction heating to densify the connecting layer, so as to densely connecting the SiC end plug to the SiC cladding tube to form a SiC cladding. In the present invention, the use of induction heating to realize quick connection between the end plug and the cladding tube greatly saves working time and improves connecting efficiency, and conductive treatment on the high-purity non-conductive SiC cladding tube and end plug is unnecessary in induction heating, such that only the connecting layer is heated, thereby achieving high reliability.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Wu, Lixiang
Xue, Jiaxiang
Liao, Yehong
Ren, Qisen
Zhai, Jianhan
Zhang, Yongdong
Zhang, Xiansheng
Abstract
Disclosed in the present invention are a connection method for a silicon carbide (SiC) cladding for nuclear application, and a SiC cladding and an application thereof. The connection method comprises the following steps: S1, preparing a connection slurry; S2, uniformly applying the connection slurry on connection surfaces of a SiC end plug and/or a SiC cladding tube, and relatively matching the connection surfaces of the SiC end plug and the SiC cladding tube to form a connection structure; S3, heating the connection structure to 100-300°C in a protective atmosphere, keeping the temperature for 0.1-4 h, and curing the connection slurry between the SiC end plug and the SiC cladding tube to form a connection layer; and S4, performing resistance welding treatment on the cured connection structure to densify the connection layer, and connecting the SiC end plug and the SiC cladding tube densely to form the SiC cladding. According to the connection method for a SiC cladding for nuclear application in the present invention, rapid connection between the end plug and the cladding tube is achieved by using the resistance welding technology, thereby greatly saving the working time, and improving the connection efficiency; and a heat-affected zone of resistance welding connection is small, such that a nuclear fuel inside the cladding cannot be affected, and the reliability of the cladding is improved.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
Inventor
Lin, Jiming
Zhao, Yuan
Qiu, Ruoxiang
Wei, Huanyi
Zeng, Xian
Zou, Qing
Duan, Chengjie
Abstract
The present invention relates to a fiber composite aerogel material, a preparation method therefor, and the use thereof. The preparation method for the fiber composite aerogel material comprises the following steps: hydrolyzing part of a silicon source to prepare a silicon hydrolysate, mixing the silicon hydrolysate with a polymer, and carrying out electrostatic spinning to prepare a nanofiber membrane; thermally treating the nanofiber membrane to remove the polymer; mixing the remaining silicon source with water and an alcohol solvent to prepare silica sol; soaking the thermally treated nanofiber membrane in silica sol, and mixing same under a negative pressure condition to prepare a nanofiber sol; and carrying out a gel aging treatment on the nanofiber sol, and then drying same to prepare the fiber composite aerogel material. By means of the preparation method for the fiber composite aerogel material, the thermal insulation property and flexibility of the fiber composite aerogel material can be improved without powder falling off.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
Inventor
Zhao, Yuan
Wei, Huanyi
Qiu, Ruoxiang
Zeng, Xian
Zou, Qing
Duan, Chengjie
Lin, Jiming
Abstract
The present invention relates to an aerogel coating and a preparation method therefor, and an aerogel coating layer and a preparation method therefor. The method for preparing the aerogel coating comprises the following steps: firstly mixing a slag powder, fly ash and sodium silicate at a mass ratio of 4 : (0.2-4) : (0.5-2) to prepare a dry powder, and then mixing the dry powder with water, wherein the mass ratio of the dry powder to water is 1 : (0.2-0.5), in order to prepare a geopolymer slurry; and mixing an aerogel powder with the geopolymer slurry to prepare the aerogel coating. The method for preparing the aerogel coating can allow the prepared aerogel coating to have a good cohesiveness, high temperature resistance and heat preservation performance.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Chen, Zhao
Shi, Kangli
Duan, Chengjie
Cui, Dawei
Qiu, Ruoxiang
Shi, Xiuan
Lin, Jiming
Abstract
The present application relates to a nuclear power supply device, comprising a graphite base body, a plurality of fuel rods, a plurality of heat pipes, and a thermoelectric converter. The graphite base body is provided with a plurality of first through holes and a plurality of second through holes; the first through holes are filled with a liquid coolant; the fuel rods are respectively placed the first through holes and are used for generating heat energy; the liquid coolant passes the generated heat energy to the graphite base body; the heat energy on the graphite base body is then transferred to the thermoelectric converter by the heat pipes placed in the second through holes for thermoelectric conversion. The liquid coolant achieves a buffering effect between the fuel rods and the graphite base body, and achieves a transition effect for the heat energy transferred to the graphite base body from the fuel rods, thereby avoiding impact of the temperature gradient and swelling of the fuel rods on the graphite base body, and improving the safety of a nuclear reaction process.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD. (China)
Inventor
Xue, Jiaxiang
Liao, Yehong
Ren, Qisen
Liu, Tong
Zhai, Jianhan
Ma, Haibin
Zhang, Xiansheng
Zhang, Yongdong
Li, Rui
Liu, Yang
Abstract
A brazing connection method for a silicon carbide cladding, comprising the following steps: S1, using Al and Si as connection materials, providing the connection materials between a cladding pipe and an end plug which are matched with each other, forming a middle connection material of an Al/Si/Al three-layer structure, and forming, by means of the middle connection material, the cladding pipe, and the end plug, an assembly to be connected; and S2, placing said assembly in a vacuum or inert atmosphere for brazing connection, the middle connection material forming a connection layer, and connecting the cladding pipe and the end plug into a whole. The present invention further relates to a silicon carbide cladding formed by the method, a fuel rod comprising same, and a fuel assembly. In the present brazing connection method, Al and Si are used as the connection materials, and the connection layer of the Al/Si/Al three-layer structure is formed between the cladding pipe and the end plug, such that high-strength and reliable connection of the SiC cladding is achieved, and the SiC cladding is good in high-temperature resistance and hydrothermal corrosion resistance, thereby meeting nuclear application requirements.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Xue, Jiaxiang
Liao, Yehong
Ren, Qisen
Liu, Tong
Zhai, Jianhan
Ma, Haibin
Zhang, Xiansheng
Zhang, Yongdong
Li, Rui
Liu, Yang
Abstract
Disclosed are a solder for connection and a preparation method therefor and a method for connecting a silicon carbide cladding. The solder for connection comprises the following raw materials: a precursor, a glass powder and an organic solvent, wherein the mass ratio of the precursor to the glass powder is 90-98: 2-10. The solder for connection of the present invention is used for connecting a silicon carbide cladding. A glass additive phase formed by the glass powder has good wettability to silicon carbide and a high connection strength. The proportion of the glass powder is adjustable, so that the coefficient of thermal expansion of the glass additive phase is adjustable, and the stress of a joint after connection is controllable. A thick compact connection layer can be realized, facilitating engineering assembly and realizing better air tightness.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Zhou, Guofeng
Zhang, Mengjin
Yu, Bing
Guan, Xuedan
Li, Wanyao
Ding, Huimin
Wu, Fengqi
Wu, Yu
Liu, Gang
Huang, Biyi
Abstract
A flexible abutment mechanism, being used for abutting two workpieces, and comprising an upper compartment (10), a lower compartment (20) connected to the upper compartment (10), a sealed connection structure (30) sealingly connecting the upper compartment (10) and the lower compartment (20), an elastic cushioning apparatus (50) provided between the upper compartment (10) and the lower compartment (20), and a displacement detection assembly (40) provided between the upper compartment (10) and the lower compartment (20) to detect a displacement between the upper compartment (10) and the lower compartment (20) when the flexible abutment mechanism performs abutment work. The displacement detection assembly (40) comprises a plurality of displacement detection units arranged at intervals in the circumferential direction of the upper compartment (10) and the lower compartment (20) and connected to an external movable abutment apparatus; when the displacement between the upper compartment (10) and the lower compartment (20) detected by each displacement detection unit is equivalent, the two workpieces are abutted in place. The flexible abutment mechanism has the advantages of a simple structure, high safety, high reliability, high abutment accuracy, etc.
B25B 25/00 - Implements for fastening, connecting, or tensioning of wire or strip
B25B 27/00 - Hand tools or bench devices, specially adapted for fitting together or separating parts or objects whether or not involving some deformation, not otherwise provided for
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Zhou, Guofeng
Zhang, Mengjin
Yu, Bing
Guan, Xuedan
Huang, Biyi
Wang, Zuhui
Lin, Xinpeng
Wu, Fengqi
Liu, Gang
Huang, Haihua
Wu, Yu
He, Zhi
Zhang, Kui
Abstract
Disclosed are a refueling transfer device and a refueling method. The refueling transfer device comprises a positioning cock (20) which is detachably arranged at an open end of a reactor core apparatus (2) to shield and seal the reactor core device (2) during refueling, a refueling transfer machine (50) which is detachably arranged on the positioning cock (20) and is capable of performing a horizontal motion and a lifting motion to take and place a fuel assembly (3), and a flexible abutting mechanism (60) which is arranged at one end of the refueling transfer machine (50). The positioning cock (20) is provided with a channel for taking and placing the fuel assembly (3); the positioning cock (20) is provided with a first isolation valve (30) arranged corresponding to the channel; a second isolation valve (40) is arranged on the side of the flexible abutting mechanism (60) facing away from the refueling transfer machine (50), and the second isolation valve (40) and the first isolation valve (30) are in abutment joint by means of the flexible abutting mechanism (60) under driving of the refueling transfer machine (50), so as to perform sealing and shielding during refueling. The refueling transfer device has high abutment joint precision, and can realize remote and automatic operation, such that the economical efficiency, safety and reliability of a reactor are improved.
G21C 19/10 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements
G21C 19/11 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements with revolving coupling elements, e.g. socket coupling
21.
METHOD AND DEVICE FOR MEASURING CONTROL ROD WORTH IN SUBCRITICAL STATE OF NUCLEAR POWER PLANT
GUANGDONG NUCLEAR POWER JOINT VENTURE CO., LTD. (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
DAYA BAY NUCLEAR POWER OPERATIONS AND MANAGEMENT CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CHINA GUANGDONG NUCLEAR POWER CO., LTD. (China)
Inventor
Tan, Shijie
Abstract
A method and a device for measuring a control rod worth in a subcritical state of a nuclear power plant. The method comprises: acquiring a first neutron detector response value of a nuclear reactor in an initial state (S201); acquiring a second neutron detector response value of the nuclear reactor in a subcritical state (S202); and determining a control rod worth according to the first neutron detector response value and the second neutron detector response value (S203). The method and the device can measure the control rod worth without occupying a power detector, and the power detector can still play a protective role on the nuclear reactor, thereby improving the security of the nuclear reactor.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD. (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD. (China)
LING DONG NUCLEAR POWER CO., LTD. (China)
Inventor
Chen, Zhao
Song, Lei
Qiu, Ruoxiang
Duan, Chengjie
Cui, Dawei
Zhang, Yong
Shi, Xiuan
Lin, Jiming
Abstract
Disclosed is a pool-type lead-based fast reactor with labyrinth-type flow channels. The pool-type lead-based fast reactor comprises a reactor vessel (1), a reactor top cover (2), a lower seal head (3), a reactor core mounting member (4), a reactor core (5), a control rod and a driving mechanism (6) thereof, a cold-hot pool partition plate (7), a first auxiliary partition plate (8), a heat exchanger (9), and a driving pump (10), wherein the reactor vessel (1) respectively forms sealing connections with the reactor top cover (2) and the lower seal head (3); the reactor core (5) is mounted at a mounting position of the reactor core mounting member (4); the control rod and the driving mechanism (6) thereof enter the reactor vessel (1) and are connected to the top of the reactor core (5); the reactor vessel (1) is partitioned, by the cold-hot pool partition plate (7), into a hot pool (T1) and a cold pool (T2) that are not in communication with each other; the first auxiliary partition plate (8) is located at an outer side of the cold-hot pool partition plate (7) and is at a certain distance from the cold-hot pool partition plate (7); the heat exchanger (9) runs through the hot pool (T1) into the cold pool (T2) to realize heat conduction; and the driving pump (10) enters the cold pool (T2) to circulate a coolant in the cold pool (T2). By means of the pool-type lead-based fast reactor with the labyrinth-type flow channels, not only can the complexity of members in the reactor and the flow resistance of a loop coolant be reduced, but the risk of tiny bubbles or water droplets entering the reactor core (5) when a heat transfer pipe of the heat exchanger (9) is broken can also be avoided.
G21C 15/12 - 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 pressure vesselArrangement 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 containment vessel
23.
PONTOON TYPE TUNNEL SHUTTLE PLATFORM, MONITORING DEVICE AND FOREIGN MATTER MONITORING METHOD
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
TAISHAN NUCLEAR POWER JOINT VENTURE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Ba, Jinyu
Wang, Guohe
Chen, Shaonan
Zhang, Meiling
Deng, Zhiyan
Yu, Bing
Wang, Huagang
Zhao, Apeng
Li, Bing
Wu, Yu
Chen, Zhi
Wu, Fengqi
Zhao, Hongji
Yin, Yong
Li, Ke
Li, Haiyang
Abstract
A pontoon type tunnel shuttle platform (30), comprising a control device (301), a first winch (302), a second winch (303), a cable (304), and a carrying platform (305); the first winch (302) and the second winch (303) are correspondingly disposed on two shield wells (10) and are both electrically connected to the control device (301); one end of the cable (304) is wound on the first winch (302), and the other end of the cable (304) successively passes through the shield well (10) under the first winch (302), a tunnel (20) and the shield well (10) under the second winch (303) and then is wound on the second winch (303); the cable (304) is located in the tunnel (20) in a tight manner and is fixedly connected with the carrying platform (305) in a sealed manner; the carrying platform (305) is electrically connected to the control device (301) by means of the tight cable (304) and is suspended in the tunnel (20); the control device (301) controls the releasing and rewinding of the cable (304) by means of the first winch (302) and the second winch (303) so that the carrying platform (305) shuttles along the cable (304) in the tunnel (20). The carrying platform (305) which can shuttle in the tunnel (20) at a stable attitude, a tunnel monitoring device (100) and a foreign matter monitoring method can be installed by a monitoring agency and can monitor the foreign matters in the tunnel (20) all the time.
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
TAISHAN NUCLEAR POWER JOINT VENTURE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Chen, Shaonan
Wang, Guohe
Ba, Jinyu
Zhang, Meiling
Liu, Shuai
Li, Bing
Ye, Huan
Li, Ke
Zhou, Jun
Hu, Haixiang
Wang, Xuezhu
Chen, Zhi
Wu, Yu
Zhao, Apeng
Wu, Fengqi
Yin, Yong
Li, Haiyang
Abstract
A tunnel wall-adhering shuttle platform (30) and a tunnel monitoring apparatus (100), the tunnel wall-adhering shuttle platform (30) being suited to shuttling in water in a tunnel (20) below a shield well (10), the tunnel wall-adhering shuttle platform (30) comprising a control apparatus (303), a winch (304), a cable (302), and a carrying platform (301) capable of floating on water, the control apparatus (303) being electrically connected to the winch (304), the carrying platform (301) being electrically connected to the control apparatus (303) by means of the cable (302), one end of the cable (302) being connected to the carrying platform (301) and the other end of the cable (302) being wound onto the winch (304), and the control apparatus (303) controlling the release and retraction of the cable (302) such that the carrying platform (301) can shuttle in a reciprocating motion along the inner wall of the tunnel (20) in a wall-adhering manner. By means of the control apparatus (303) controlling the release and retraction of the cable (302), the carrying platform (301) can shuttle in a reciprocating motion and maintain a stable attitude on the inner wall of the tunnel (20), and a monitoring mechanism (40) can be mounted on the carrying platform (301) in order to monitor the condition of foreign matter throughout the tunnel (20).
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
TAISHAN NUCLEAR POWER JOINT VENTURE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Wang, Guohe
Deng, Zhiyan
Chen, Shaonan
Liu, Shuai
Zhang, Meiling
Wu, Yu
Li, Bing
Ba, Jinyu
Yu, Bing
Dong, Pengfei
Wu, Fengqi
Zhao, Apeng
Chen, Zhi
Yin, Yong
Li, Ke
Li, Haiyang
Abstract
A cleaning and collecting system for a tunnel, comprising a cleaning robot (200), a collecting robot (100) and a remote operating apparatus (300). The cleaning robot (200) comprises a moving carrier (5), a mechanical arm assembly, an end cleaning apparatus (7) and a local control system, wherein the local control system comprises a local controller (81), a first sensing system (82) and a first camera system (83), the local controller (81) controlling the moving carrier (5), the mechanical arm assembly and the end cleaning apparatus (7) according to information from the first sensing system (82). The collecting robot (100) comprises a moving vehicle body (1), a stripping and conveying apparatus (2), a collecting box (3) and a local control system, wherein the local control system comprises a local controller (41), a second sensing system (42) and a second camera system (43), the local controller (41) controlling the moving vehicle body (1) and the stripping and conveying apparatus (2) according to information from the second sensing system (42). The remote operating apparatus (300) receives information from the local controller (81) and the local controller (41) and from the first camera system (83) and the second camera system (43), and controls the moving carrier (5) and the mechanical arm assembly by means of the local controller (81) and controls the moving vehicle body (1) and the stripping and conveying apparatus (2) by means of the local controller (41).
CHINA NUCLEAR POWER TECHNOLOGY RESEARCH INSTITUTE CO., LTD (China)
CHINA GENERAL NUCLEAR POWER CORPORATION (China)
CGN POWER CO., LTD (China)
Inventor
Li, Weicai
Yu, Wenchi
Fu, Xiangang
Zhou, Yuemin
Zhang, Guoliang
Chen, Jianxin
Abstract
Provided are a fuel assembly and a method for assembling same. The fuel assembly comprises a lattice frame and fuel rods (14) mounted in the lattice frame in a longitudinal direction, wherein the lattice frame comprises a plurality of strips (131), the strips (131) are intersected and matched with each other to form a plurality of grid units (130), the fuel rods (14) are correspondingly located inside the grid units (130), the strip (131) is further provided with a blending wing (16), the strip (131) and the blending wing (16) are of an integral structure formed by punching, and the blending wing (16) further comprises a secondary forming part. Since the secondary forming preparation precision has been improved, influences caused by a dimensional tolerance and an assembling tolerance are reduced, thereby enhancing the generated transverse flow, and improving efficacy of a coolant fluid. Moreover, fluid layers on the surfaces of the fuel rods (14) are effectively destroyed in the vicinity of the fuel rods (14), enhancing the blending effect.