CN203658188U - Slight impact wear test device for heat transfer tube of steam generator of nuclear power plant - Google Patents
Slight impact wear test device for heat transfer tube of steam generator of nuclear power plant Download PDFInfo
- Publication number
- CN203658188U CN203658188U CN201320760268.6U CN201320760268U CN203658188U CN 203658188 U CN203658188 U CN 203658188U CN 201320760268 U CN201320760268 U CN 201320760268U CN 203658188 U CN203658188 U CN 203658188U
- Authority
- CN
- China
- Prior art keywords
- steam generator
- transfer pipe
- nuclear power
- heat
- heat transfer
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Withdrawn - After Issue
Links
Images
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
The utility model discloses a slight impact wear test device for a heat transfer tube of a steam generator of a nuclear power plant. The slight impact wear test device consists of an electromagnetic vibration excitation driving device, a leveling lifting device, an environment control device, a data acquisition control system and a frame, wherein the main structure of the electromagnetic vibration excitation driving device is as follows: an electromagnetic vibration exciter is hung below a middle beam of the frame, the end part of the electromagnetic vibration exciter is connected with an upper fixture for clamping a block-shaped upper test piece through a force transducer, one end of a non-contact type displacement sensor is fixed on the upper fixture, and the other end of the non-contact type displacement sensor is fixed on the vibration exciter; the structure of the leveling lifting device is as follows: a V-shaped groove lower fixture for clamping the heat transfer tube of the steam generator is fixed on an electric heating plate, and the bottom of the electric heating plate is connected with a base through a heat insulation plate, a lifting platform, a leveling platform and a transition base in sequence. The test device disclosed by the utility model can simulate the slight impact wear of the heat transfer tube of the steam generator of the nuclear power plant more effectively so as to provide an accurate and reliable test basis for the design, manufacturing and maintenance of the heat transfer tube of the steam generator of the nuclear power plant.
Description
Technical field
The utility model relates to a kind of impact wear testing machine a little of nuclear power station steam generator heat-transfer pipe.
Background technology
Impact wear is a solid surface and another solid surface surface abrasion that dynamic Contact, shock cause repeatedly.In all kinds of wear forms, impact wear is owing to there being impulsive force, thereby the most serious to the damage of material.In modern industry, heat interchanger is very important unit.Its kind is a lot, and wherein tubular heat exchanger (heat-transfer pipe) is most widely used one.This class heat interchanger is that heat-transfer pipe exists a serious shortcoming: the power machine vibration that the flow-induced vibration in pipe and its connecting piece are propagated is easily vibrated it, forms impact wear and then breakage.
Two planar part surface in contacts of the steam generator heat-transfer pipe of nuclear power station and anti-vibration strip or back up pad seem motionless; But in fact, under flow-induced effect of vibration, between anti-vibration strip or back up pad and heat-transfer pipe, can there is the vibration of the invisible small amplitude of naked eyes, form impact wear a little.Because the wall thickness of this kind of heat-transfer pipe is only 1mm left and right, impact wear easily causes its breakage a little.Because steam generator heat-transfer pipe is the visual plant of nuclear power plant, it is a loop equipment, is again secondary circuit equipment, is a loop and secondary circuit binding site, bears one, secondary circuit energy transmission tasks; Both in the face of high temperature, the pressurized working fluid condition in a loop, faced again the severe condition of work of secondary circuit water quality.Its a little the environment of impact wear be not air drying environment, the a little impact wear of heat-transfer pipe under this rugged surroundings more easily causes its leakage and breaks, and heat-transfer pipe One's name is legion is (as the steam generator of the electric power nuclear power station that is 300MW, 5000 heat-transfer pipes of just having an appointment), therefore, heat-transfer pipe is the nuclear power station even main cause of accident that breaks down because impact wear a little causes breaking.
The impact wear a little of nuclear power station steam generator heat-transfer pipe is tested and analyzed, to get its abrasion mechanism and the relation to relevant operating mode clear, can provide accurately for the design of nuclear power station steam generator heat-transfer pipe, manufacture and maintenance, reliable test basis, with reduce crackle germinating, expand and break, improve the fatigue lifetime of heat-transfer pipe, and then improve nuclear power station reliability of operation and security.But at present, there is no special is nuclear power heat-transfer pipe experimental assembly and the test method of the design of impact wear experimental amount body a little, without heat-transfer pipe impact wear infrastest data accumulation, corresponding abrasion mechanism model and Life Prediction Model a little, this will restrict integrity assessment and life prediction to steam generator, and then affects nuclear plant safety, reliability service.
Utility model content
The purpose of this utility model is to provide a kind of nuclear power station steam generator heat-transfer pipe impact abrasion testing equipment a little, this test unit can be simulated the impact wear a little between nuclear power station steam generator heat-transfer pipe and anti-vibration strip or back up pad, analyze with the impact wear a little to steam generator heat-transfer pipe, get its abrasion mechanism and the relation to relevant operating mode clear, thereby be the design of nuclear power station steam generator heat-transfer pipe, manufacture with maintenance and provide accurate, test basis reliably, to reduce the germinating of its crackle, expand and break, improve the fatigue lifetime of heat-transfer pipe, and then raising nuclear power station reliability of operation and security.
The utility model is realized the technical scheme that its goal of the invention adopts: a kind of nuclear power station steam generator heat-transfer pipe impact abrasion testing equipment a little, it is made up of electromagnetic exciting drive unit, leveling jacking gear, environment control unit, data acquisition control system and frame, wherein:
The formation of described frame is: on base, be provided with portal frame, on two columns of portal frame, be sheathed with can vertical movement central sill, the upper surface at central sill middle part is connected with the back timber of portal frame by screw pair;
Consisting of of electromagnetic exciting drive unit: electromagnetic exciter by hung bolt under central sill, the end of electromagnetic exciter is connected with the upper fixture of the block upper test specimen of clamping by force cell, and the vertical center line in the clamping chamber of upper fixture and the axle centering of electromagnetic exciter; One end of non-contact displacement transducer is fixed on fixture, the other end is fixed on vibrator;
Consisting of of leveling jacking gear: " V " type groove lower clamp of clamping steam generator heat-transfer pipe is fixed on electric hot plate, electric hot plate bottom is connected with transition base by thermal insulation board, lifting table, leveling platform successively; The T shape guide rail of leveling platform bottom coordinates with the T-slot of base;
Consisting of of environment control unit: be fixed with the water cavity that surrounds test specimen and " V " type groove lower clamp under cylindrical tube on electric hot plate, the electric hot plate upper surface that water cavity surrounds is provided with temperature sensor; Temperature sensor, electric hot plate are all electrically connected with outside temperature controller.
Electromagnetic exciter, displacement transducer and force cell are all electrically connected with data acquisition control system.
A kind of principle of work of the present utility model is:
A, steam generator heat-transfer pipe is clamped on " V " type groove lower clamp, block test specimen is clamped on upper fixture; Regulate the vertical height of central sill by screw pair, make block test specimen and heat-transfer pipe close, the inclination angle, front and back of adjusting again leveling platform makes the axis of heat-transfer pipe and the plane keeping parallelism of block test specimen, regulate subsequently the height of lifting table to make block test specimen and heat-transfer pipe keep good line to contact, and under the retroactive effect of force cell, apply the initial vertical contact load P of setting to steam generator heat-transfer pipe;
Temperature controller, under the retroactive effect of temperature sensor, makes electric hot plate reach design temperature simultaneously;
The motion of b, data acquisition control system control electromagnetic exciter, the block test specimen that makes fixture and clamping thereof is with the frequency set and multiplicity is vertical moves back and forth, and realizes the impact wear a little under design temperature environment between the steam generator heat-transfer pipe that block test specimen and " V " type groove lower clamp clamp; Meanwhile, connected displacement transducer is measured the displacement amplitude of block test specimen, and delivers to data acquisition control system.
Another kind of principle of work of the present utility model is:
A, steam generator heat-transfer pipe is clamped on " V " type groove lower clamp, block test specimen is clamped on upper fixture; Regulate the vertical height of central sill by screw pair, make block test specimen and heat-transfer pipe close, the inclination angle, front and back of adjusting again leveling platform makes the axis of heat-transfer pipe and the plane keeping parallelism of block test specimen, regulate subsequently the height of lifting table to make block test specimen and heat-transfer pipe keep good line to contact, and under the retroactive effect of force cell, apply the initial vertical contact load P of setting to steam generator heat-transfer pipe;
Meanwhile, be filled with liquid medium to water cavity, make liquid medium flood the contact site of heat-transfer pipe and block test specimen, temperature controller, under the retroactive effect of temperature sensor, makes electric hot plate reach design temperature;
The motion of b, data acquisition control system control electromagnetic exciter, the block test specimen that makes fixture and clamping thereof is with the frequency set and multiplicity is vertical moves back and forth, and realizes the impact wear a little under the liquid medium environment of design temperature between the steam generator heat-transfer pipe that block test specimen and " V " type groove lower clamp clamp; Meanwhile, connected displacement transducer is measured the displacement amplitude of block test specimen, and delivers to data acquisition control system.
Compared with prior art, the beneficial effects of the utility model are:
One, the utility model is connected with vertical electromagnetic exciter block test specimen respectively with steam generator heat-transfer pipe with the leveling jacking gear that is positioned at base, and the temperature environment of the setting reaching by the heating of electric hot plate, drive block test specimen to do radially (vertical) motion by electromagnetic exciter again, thereby realize under specified temp environment the simulation of the ballistic motion a little between the steam generator heat-transfer pipe to nuclear power station and anti-vibration strip or back up pad.Displacement, frequency and temperature in data acquisition control system impact motion process, vertical impact load record and storage, in conjunction with observation and the record of state of wear; Can analyze the impact wear a little of nuclear power station steam generator heat-transfer pipe, to get its abrasion mechanism and the relation to relevant operating mode clear, thereby for design, manufacture and the maintenance of nuclear power station steam generator heat-transfer pipe provide accurately, reliable test basis, with reduce crackle germinating, expand and break, improve fatigue lifetime of heat-transfer pipe.
Two, the start amplitude of electromagnetic exciter is controlled, and frequency is high, and precision is high, thereby can realize the impact wear test under accurately imposing a condition between block test specimen and heat-transfer pipe; Its test findings more accurately, reliably.
Three, by leveling and the lifting of leveling jacking gear, can make to keep good line to contact between block test specimen and heat-transfer pipe, thereby the simulation of the impact wear a little between the anti-vibration strip or the back up pad that can more real heat-transfer pipe contact with its line, also make its test findings more accurately, reliable.
In addition, test time is filled with liquid medium to water cavity, and to make the temperature of liquid medium by temperature controller by electric hot plate be design temperature; Can realize the simulation impact wear test a little between heat-transfer pipe and the block test specimen in design temperature liquid medium.The actual condition of its environmental baseline and heat-transfer pipe and anti-vibration strip or back up pad is more approaching, and its test findings is more reliable, can provide more accurate, reliable test basis for the design of nuclear power station steam generator heat-transfer pipe, manufacture and maintenance.
Below in conjunction with the drawings and specific embodiments, the utility model is described in further detail.
Brief description of the drawings
Fig. 1 is the main TV structure schematic diagram of the utility model embodiment.
Fig. 2 is the A-A sectional structure enlarged diagram of Fig. 1.
Embodiment
Embodiment
Fig. 1, Fig. 2 illustrate, a kind of embodiment of the present utility model is, a kind of nuclear power station steam generator heat-transfer pipe impact abrasion testing equipment a little, it is made up of electromagnetic exciting drive unit, leveling jacking gear, environment control unit, data acquisition control system and frame, wherein:
The formation of described frame is: on base 10, be provided with portal frame, on two columns 11 of portal frame, be sheathed with can vertical movement central sill 12, the upper surface at central sill 12 middle parts is connected with the back timber 13 of portal frame by screw pair 14;
Consisting of of electromagnetic exciting drive unit: electromagnetic exciter 21 by hung bolt central sill 12 times, the end of electromagnetic exciter 21 is connected with the upper fixture 23 of the block upper test specimen 25 of clamping by force cell 22, and the vertical center line in the clamping chamber of upper fixture 23 and the axle centering of electromagnetic exciter 21; One end of non-contact displacement transducer 24 is fixed on fixture 23, the other end is fixed on vibrator 21;
Consisting of of leveling jacking gear: " V " type groove lower clamp 35 of clamping steam generator heat-transfer pipe 38 is fixed on electric hot plate 36, and electric hot plate 36 bottoms are connected with transition base 31 by thermal insulation board 34, lifting table 33, leveling platform 32 successively; The T shape guide rail of leveling platform 32 bottoms coordinates with the T-slot of base 10;
Consisting of of environment control unit: be fixed with the water cavity 39 that surrounds test specimen 38 and " V " type groove lower clamp 35 under cylindrical tube on electric hot plate 36, electric hot plate 36 upper surfaces that water cavity 39 surrounds are provided with temperature sensor 37; Temperature sensor 37, electric hot plate 36 are all electrically connected with outside temperature controller.
Use this routine nuclear power station steam generator heat-transfer pipe impact abrasion testing equipment a little, carry out steam generator heat-transfer pipe a kind of test method of impact wear a little, its practice is:
A, by steam generator heat-transfer pipe 38 be clamped on " V " type groove lower clamp 35, block test specimen 25 is clamped on upper fixture 23; Regulate the vertical height of central sill 12 by screw pair 14, make block test specimen 25 close with heat-transfer pipe 38, the inclination angle, front and back of adjusting again leveling platform 32 makes the axis of heat-transfer pipe 38 and the plane keeping parallelism of block test specimen 25, regulate subsequently the height of lifting table 33 to make block test specimen 25 and heat-transfer pipe 38 keep good line to contact, and under the retroactive effect of force cell 22, apply the initial vertical contact load P of setting to steam generator heat-transfer pipe 38;
Temperature controller, under the retroactive effect of temperature sensor 37, makes electric hot plate reach design temperature simultaneously;
The motion of b, data acquisition control system control electromagnetic exciter 21, the block test specimen 25 that makes fixture 23 and clamping thereof is with the frequency set and multiplicity is vertical moves back and forth, and realizes the impact wear a little under design temperature environment between the steam generator heat-transfer pipe 38 that block test specimen 25 and " V " type groove lower clamp 35 clamp; Meanwhile, connected displacement transducer 24 is measured the displacement amplitude of block test specimen, and delivers to data acquisition control system.
Use this routine nuclear power station steam generator heat-transfer pipe impact abrasion testing equipment a little, carry out the steam generator heat-transfer pipe another kind of test method of impact wear a little, its practice is:
A, by steam generator heat-transfer pipe 38 be clamped on " V " type groove lower clamp 35, block test specimen 25 is clamped on upper fixture 23; Regulate the vertical height of central sill 12 by screw pair 14, make block test specimen 25 close with heat-transfer pipe 38, the inclination angle, front and back of adjusting again leveling platform 32 makes the axis of heat-transfer pipe 38 and the plane keeping parallelism of block test specimen 25, regulate subsequently the height of lifting table 33 to make block test specimen 25 and heat-transfer pipe 38 keep good line to contact, and under the retroactive effect of force cell 22, apply the initial vertical contact load P of setting to steam generator heat-transfer pipe 38;
Meanwhile, be filled with liquid medium to water cavity 36, make liquid medium flood the contact site of heat-transfer pipe 38 and block test specimen 25, temperature controller, under the retroactive effect of temperature sensor 37, makes electric hot plate reach design temperature;
The motion of b, data acquisition control system control electromagnetic exciter 21, the block test specimen 25 that makes fixture 23 and clamping thereof is with the frequency set and multiplicity is vertical moves back and forth, and realizes the impact wear a little under the liquid medium environment of design temperature between the steam generator heat-transfer pipe 38 that block test specimen 25 and " V " type groove lower clamp 35 clamp; Meanwhile, connected displacement transducer 24 is measured the displacement amplitude of block test specimen, and delivers to data acquisition control system.
Claims (1)
1. a nuclear power station steam generator heat-transfer pipe impact abrasion testing equipment a little, it is made up of electromagnetic exciting drive unit, leveling jacking gear, environment control unit, data acquisition control system and frame, wherein:
The formation of described frame is: base is provided with portal frame on (10), on two columns (11) of portal frame, be sheathed with can vertical movement central sill (12), the upper surface at central sill (12) middle part is connected with the back timber (13) of portal frame by screw pair (14);
Consisting of of electromagnetic exciting drive unit: electromagnetic exciter (21) by hung bolt under central sill (12), the end of electromagnetic exciter (21) is connected with the upper fixture (23) of the block upper test specimen (25) of clamping by force cell (22), and the axle centering of the vertical center line in the clamping chamber of upper fixture (23) and electromagnetic exciter (21); One end of non-contact displacement transducer (24) is fixed on fixture (23), the other end is fixed on vibrator (21);
Consisting of of leveling jacking gear: it is upper that " V " type groove lower clamp (35) of clamping steam generator heat-transfer pipe (38) is fixed on electric hot plate (36), electric hot plate (36) bottom is connected with transition base (31) by thermal insulation board (34), lifting table (33), leveling platform (32) successively; The T shape guide rail of leveling platform (32) bottom coordinates with the T-slot of base (10);
Consisting of of environment control unit: be fixed with the water cavity (39) that surrounds test specimen under cylindrical tube (38) and " V " type groove lower clamp (35) on electric hot plate (36), electric hot plate (36) upper surface that water cavity (39) surrounds is provided with temperature sensor (37); Temperature sensor (37), electric hot plate (36) are all electrically connected with outside temperature controller;
Electromagnetic exciter (21), displacement transducer (24) and force cell (22) are all electrically connected with data acquisition control system.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201320760268.6U CN203658188U (en) | 2013-11-27 | 2013-11-27 | Slight impact wear test device for heat transfer tube of steam generator of nuclear power plant |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201320760268.6U CN203658188U (en) | 2013-11-27 | 2013-11-27 | Slight impact wear test device for heat transfer tube of steam generator of nuclear power plant |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CN203658188U true CN203658188U (en) | 2014-06-18 |
Family
ID=50924565
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN201320760268.6U Withdrawn - After Issue CN203658188U (en) | 2013-11-27 | 2013-11-27 | Slight impact wear test device for heat transfer tube of steam generator of nuclear power plant |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN203658188U (en) |
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN103592195A (en) * | 2013-11-27 | 2014-02-19 | 西南交通大学 | Machine and method for slight impact wear test on nuclear power station steam generator heat-transfer pipe |
| CN104373427A (en) * | 2014-09-11 | 2015-02-25 | 常州市武进牛塘红星制刷厂 | Rod fixing device |
| CN110047601A (en) * | 2019-05-22 | 2019-07-23 | 上海核工程研究设计院有限公司 | A kind of U-shaped heat-transfer pipe single tube dynamic property tester of nuclear power plant's vertical steam generator |
-
2013
- 2013-11-27 CN CN201320760268.6U patent/CN203658188U/en not_active Withdrawn - After Issue
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN103592195A (en) * | 2013-11-27 | 2014-02-19 | 西南交通大学 | Machine and method for slight impact wear test on nuclear power station steam generator heat-transfer pipe |
| CN103592195B (en) * | 2013-11-27 | 2016-08-17 | 西南交通大学 | Nuclear power station steam generator heat-transfer pipe impact wear testing machine a little and test method |
| CN104373427A (en) * | 2014-09-11 | 2015-02-25 | 常州市武进牛塘红星制刷厂 | Rod fixing device |
| CN110047601A (en) * | 2019-05-22 | 2019-07-23 | 上海核工程研究设计院有限公司 | A kind of U-shaped heat-transfer pipe single tube dynamic property tester of nuclear power plant's vertical steam generator |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN103592195A (en) | Machine and method for slight impact wear test on nuclear power station steam generator heat-transfer pipe | |
| CN103604713A (en) | Multidirectional fretting wear device and testing method for heat transmission pipe of steam generator | |
| US11740167B2 (en) | Temperature-controllable large-size geotechnique true triaxial multi-field coupling test system and test method | |
| CN110595909B (en) | True triaxial test system and method for simulating deep rock mass under different temperature influences | |
| CN210775067U (en) | True triaxial test system for simulating different temperature influences of deep rock mass | |
| CN103454138B (en) | Horizontal geotechnical plane stress triaxial apparatus for pressure chamber structure | |
| CN104677815A (en) | True triaxial rock parameter test system | |
| CN109540661B (en) | Dynamic stress and gradient stress combined loading experimental device and method | |
| CN203658188U (en) | Slight impact wear test device for heat transfer tube of steam generator of nuclear power plant | |
| CN109580399A (en) | Middle low strain dynamic rate sound integration test system | |
| CN203551383U (en) | In-situ testing instrument for micromechanical property of material under multi-load and multi-physical field coupling action | |
| CN204142594U (en) | A kind of high-temperature high-pressure in-situ fretting wear test unit | |
| CN109001040B (en) | Rock fracturing simulator | |
| CN110687272A (en) | Rock joint face shear seepage test device and test method based on geothermal exploitation | |
| CN106289962B (en) | In-situ test system capable of observing deformation and damage of sample gauge length section in high-low power on-line manner | |
| CN104483046B (en) | One kind surveys shear device and measuring method | |
| CN102494962A (en) | Circulating single shaft dynamic load testing device with function of simulating tire load | |
| CN203658187U (en) | Multidirectional fretting wear equipment for heat transfer tubes of steam generator | |
| CN106950137B (en) | Tangential fretting wear test device and test method | |
| CN207622905U (en) | A kind of test device of ultrasonic activation detritus Effective power | |
| CN204142199U (en) | A kind of portable large curved surface on-line measurement and matching device | |
| CN102252643B (en) | Solar thermal generation reflector lens curved surface testing system | |
| CN203069469U (en) | Tangential and radial composite fretting corrosive abrasion test equipment | |
| CN203376226U (en) | Solar vacuum tube cold and hot impact test device | |
| KR20150104494A (en) | Device for Calculation of Physical Property of Model Ice in Ice Basin |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| C14 | Grant of patent or utility model | ||
| GR01 | Patent grant | ||
| AV01 | Patent right actively abandoned |
Granted publication date: 20140618 Effective date of abandoning: 20160817 |
|
| C25 | Abandonment of patent right or utility model to avoid double patenting |