CN104425045B - One loop of nuclear power station pressurizer system information processing method and investigation method - Google Patents
One loop of nuclear power station pressurizer system information processing method and investigation method Download PDFInfo
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- CN104425045B CN104425045B CN201310413734.8A CN201310413734A CN104425045B CN 104425045 B CN104425045 B CN 104425045B CN 201310413734 A CN201310413734 A CN 201310413734A CN 104425045 B CN104425045 B CN 104425045B
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- G—PHYSICS
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- G21D—NUCLEAR POWER PLANT
- G21D3/00—Control of nuclear power plant
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Abstract
The application is related to one loop of nuclear power station pressurizer system information processing method and investigation method, including：According to the structural relation of pressure release case liquid level during normal operation and pressure release case, the corresponding relation of pressure release case liquid level rise speed and primary Ioops slip is determined, the unit identical numerical value with primary Ioops slip is converted to current pressure release case liquid level rise speed according to corresponding relation；According to corresponding relation, the matching relationship that pressure release case liquid level rise speed handles prediction scheme with primary Ioops slip is set up.The application is by the way that liquid level rise speed to be converted to the primary Ioops slip of existing standard, the long-standing pressure release case liquid level of nuclear power station is solved to go up without typical problem, the matching relationship of prediction scheme is handled with primary Ioops slip by the liquid level rise speed of foundation, solve the problem of pressure release case liquid level existed for a long time in nuclear power station goes up without intervention means, so as to reduce unnecessary valve Disintegration overhaul, and then reduce the radioactive dose and radwaste of staff.
Description
Technical field
The application is related to technical field of nuclear power, more particularly to a kind of suitable for pressurized-water reactor nuclear power plant primary Ioops pressurizer system
Information processing method and a kind of liquid level rise speed suitable for pressurized-water reactor nuclear power plant primary Ioops voltage-stablizer pressure release case are too fast
Investigate processing method.
Background technology
Nuclear power station（Nuclear Power Plant）It is to utilize nuclear fission (Nuclear Fission) or nuclear fusion
(Nuclear Fusion) reacts the power plant of discharged energy production electric energy.It is, in principle, that nuclear power station realize nuclear energy-
The energy conversion of heat energy-electric energy.Said in terms of slave unit, the reactor and steam generator of nuclear power station are served equivalent to thermal power station
Fossil fuel and boiler effect.Energy conversion in nuclear power station is realized by means of three loops.Reactor coolant exists
Enter reactor under the driving of main pump, flow through after reactor core and to be flowed out from the outlet of reactor vessel, into steam generator, then
Main pump is returned to, here it is the circulation process of reactor coolant（Also known as primary Ioops）.During circulating, reactor is cold
But the heat of nuclear reaction generation is taken away in agent from reactor core, and in a vapor generator, passes heat under conditions of being physically isolated
Pass the water of secondary circuit.Secondary circuit water is heated, and the steam of generation removes driving steam turbine again, drives and the generating of steam turbine coaxially
Machine generates electricity.Weary steam after acting is within the condenser by seawater or river, lake-water cooling water（Three Loop Waters）Water is condensed into, then is mended
It is charged in steam generator.The effect in three loops by medium of seawater is that weary steam is condensed into water, while taking away power station
Abandon heat.
One loop of nuclear power station system is a loop, therefore, is caused when the cooling agent in system produces temperature change
During volume fluctuations, system pressure certainly will be caused to produce corresponding change.If system pressure increases to over design pressure, it will cause
System and damage of facilities；If pressure decline is too low, reactor core local boiling or bulk boiling can be caused again, causes reactor core to burn,
So as to trigger serious safety problem.Therefore in order to which nuclear plant safety is reliably run, it is necessary to reactor-loop system
Pressure is controlled and protected, here it is the purpose of pressure relief system is set up, and voltage-stablizer is then the master of pressure relief system
Want equipment.
The main function of voltage-stablizer is to maintain the pressure of primary Ioops on setting valve, to prevent coolant water in primary Ioops
Middle vaporization.When the pressure of voltage-stablizer exceedes the setting valve of safety valve, safety valve is opened, and the steam in voltage-stablizer is drained into rapidly
In release case, make voltage-stablizer release, play overpressure protection effect.
The major function of voltage-stablizer relief valve is collection, condensation and cooling pressurizer safety valve, residual heat removal system（RRA）
Safety valve, chemistry and volume system（RCV）What the positive steam and primary Ioops system valve rod filler device of safety valve discharge were leaked
Cooling agent.Pressure release case makes the cooling agent of primary Ioops not discharged to containment vessel, it is to avoid with active primary Ioops fluid
Pollution to containment.
Under normal circumstances, voltage-stablizer pressure release case liquid level rise speed should be almost unchanged, close to fixed value.But through live real
Border is it has been observed that the very fast problem of voltage-stablizer pressure release case liquid level rise speed repeatedly occurs repeatedly.However, because current nuclear power station does not have
Road voltage-stablizer pressure release case liquid level rise standard limit, does not also go up for pressure release case liquid level and provides intervention means, lead once
This problem is caused to bring greatly puzzlement to nuclear power station staff and repeatedly paid close attention to by nuclear safety office.Moreover, dynamic through U.S.'s core
Power operation study institute（INPO）Feedback, U.S.'s nuclear power station once occurred together voltage-stablizer shower valve packing leakage, cause one time
Road system leak reactor promptly stops the event of pen, and the leakage of voltage-stablizer shower valve packing is to be recycled to voltage-stablizer pressure release case, from
And cause pressure release case liquid level to go up.It is currently for the too fast common practice of pressure release case liquid level rise speed, in each overhaul
When strip inspection primary Ioops valve, the radioactive dose of nuclear power station staff is adds additional, while generating a large amount of radioactivity
Waste, and add financial cost.
The content of the invention
The application provides a kind of suitable for pressurized-water reactor nuclear power plant primary Ioops pressurizer system information processing method and one kind
The too fast investigation processing method of liquid level rise speed suitable for pressurized-water reactor nuclear power plant primary Ioops voltage-stablizer pressure release case, to solve
The problem of current one loop of nuclear power station voltage-stablizer pressure release case liquid level goes up without standard, without intervention means.
According to the application's in a first aspect, the application provides a kind of one loop of nuclear power station pressurizer system information processing side
Method, including：To should determine that step, according to the structural relation of pressure release case liquid level during normal operation and pressure release case；Obtain
Step, obtains current pressure release case liquid level rise speed；Scaling step, determines pressure release case liquid level rise speed and primary Ioops slip
Corresponding relation, the unit with primary Ioops slip is converted to current pressure release case liquid level rise speed according to the corresponding relation
Identical numerical value；Establishment step is matched, according to the alignment processing prediction scheme of the primary Ioops slip, the rise of pressure release case liquid level is set up
Speed handles the matching relationship of prediction scheme with primary Ioops slip.This method is by the way that voltage-stablizer pressure release case liquid level rise speed is changed
For the primary Ioops slip of existing standard, solve the voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station and go up without mark
Quasi- problem.
Further, the scaling step includes：The corresponding per unit slip of per unit liquid level is calculated, according to calculating
As a result and historical defect data, the corresponding relation of pressure release case liquid level rise speed and primary Ioops slip is determined；According to institute
The reading of the connected liquid level gauge of pressure release case is stated, slip corresponding with current pressure release case liquid level rise speed is calculated, it is described to let out
Leak rate is the unit identical numerical value with primary Ioops slip.
Further, the pressure release case be shaped as centre be cylinder and both sides are volume identical spherical crown, it is described just
Often the liquid level during operation is 2 meters to 2.2 meters, and the corresponding per unit slip of per unit liquid level that calculates includes：
Liquid surface volume when liquid level is 2 meters and 2.2 meters is calculated respectively, further according to spherical crown volume and cylinder volume formula, is calculated
Liquid level often go up 1 millimeter when volume rate of change, so as to obtain corresponding primary Ioops slip.Due to general one loop of nuclear power station
Shape, size and its built-in pipeline of voltage-stablizer pressure release case etc. are standard configuration, and obtained per unit liquid level correspondence is calculated accordingly
Per unit slip be generally also it is fixed, therefore, according to the reading of liquid level gauge can be easily calculated pressure release case work as
The corresponding primary Ioops slip of preceding liquid level rise speed so that nuclear power station operator soon can make according to the result of calculation
Corresponding reaction.
Further, the pressure release case liquid level rise speed and the determination step bag of the corresponding relation of primary Ioops slip
Include：If pressure release case liquid level rise speed is less than first rate designated value, it is determined that its corresponding primary Ioops slip is small
Designated value is leaked in first；If pressure release case liquid level rise speed is more than or equal to first rate designated value and less than the second speed
Rate designated value, it is determined that its corresponding primary Ioops slip is more than or equal to the first leakage designated value；If pressure release case liquid level
Rise speed is more than or equal to the second speed designated value, it is determined that its corresponding primary Ioops slip is more than or equal to second
Leak designated value；The first rate designated value and the second speed designated value are set according to historical defect data, and described first lets out
Leakage designated value is calculates obtained numerical value according to the scaling step and with reference to the first rate designated value, and described second leaks
Designated value is calculates obtained numerical value according to the scaling step and with reference to the second speed designated value.
Further, the corresponding primary Ioops slip processing prediction scheme of the first leakage designated value includes：According to normal prison
The frequency monitoring pressure release case liquid level rise situation of nuclear power station is surveyed, or monitors pressure release case liquid level rise situation and tracks pressure release case liquid level
The trend of rise；Described second, which leaks the corresponding primary Ioops slip processing prediction scheme of designated value, includes：Determine in pressure release case liquid level
Existing exception is swelled, checks and determines pressure release case liquid level rise reason.The embodiment is let out by historical defect data and primary Ioops
The related running technology specification of leak rate, can learn the liquid level rise speed of one loop of nuclear power station voltage-stablizer pressure release case in what value
Belong to normal monitoring range, therefore the correlation for the corresponding primary Ioops slip of liquid level rise speed that can be read according to liquid level gauge
Running technology specification, come determine whether to monitor as usual or strengthen monitoring, and judge pressure release case liquid level rise whether occur it is different
Often.
Preferably, the first rate designated value is 10 millimeters daily, and the first leakage designated value is 6.25 liters per small
When, the second speed designated value is 15 millimeters daily, and the second leakage designated value is 9.375 liter per hour.
According to the second aspect of the application, the application provides a kind of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed
The too fast investigation processing method of rate, including：Establishment step, it is corresponding with primary Ioops slip according to pressure release case liquid level rise speed
Relation, is converted to the unit identical numerical value with primary Ioops slip, and set up pressure release by current pressure release case liquid level rise speed
Case liquid level rise speed handles the matching relationship of prediction scheme with primary Ioops slip；Step is selected, according to numerical value, is closed in the matching
Corresponding primary Ioops slip processing prediction scheme is selected in system；Step is investigated, the primary Ioops slip processing prediction scheme selected is performed,
And when appearance is abnormal it is determined that pressure release case liquid level goes up, checks and determine pressure release case liquid level rise reason.This method is by setting up
Liquid level rise speed and primary Ioops slip handle the matching relationship of prediction scheme so that can be obtained after liquid level rise speed is obtained
Know corresponding processing prediction scheme, solve the voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station and go up without intervention means
Problem.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：Perform safety valve
Release pressure fixed value checking and air seal test, if verification and result of the test are unqualified, it is determined that pressure release case liquid level rise reason
There is interior leakage phenomenon for the spring loaded safety valve of chemical volume control system.The embodiment is determined in liquid level by safety valve testing
Whether the reason that rises is due to leak phenomenon in the spring loaded safety valve appearance of chemical volume control system and cause.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：Check that chemistry holds
The pilot operated safety valve discharge pipe line temperature of product control system, judges whether the trend of its temperature rise phenomenon occurs, if
Go up, it is determined that pressure release case liquid level rise reason is that the pilot operated safety valve of chemical volume control system interior leakage phenomenon occurs.Should
Embodiment determines whether liquid level rise reason is due to that the pilot operated safety valve of chemical volume control system goes out by temperature inspection
Phenomenon is leaked in existing and is caused.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：Check reactor
The pilot operated safety valve discharge pipe line temperature of residual heat removal system, judges whether the trend of its temperature rise phenomenon occurs, if
There is rise, it is determined that it is existing that pressure release case liquid level rise reason interior leakage occurs for the pilot operated safety valve of reactor RHR system
As.The embodiment by temperature inspection determine liquid level rise reason whether be due to reactor RHR system pilot-operated type pacify
Full valve is leaked phenomenon and caused in occurring.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：Check reactor
The pilot operated safety valve discharge pipe line temperature of coolant system, judges whether the trend of its temperature rise phenomenon occurs, if
Go up, it is determined that pressure release case liquid level rise reason interior leakage phenomenon occurs for the pilot operated safety valve of reactor coolant loop.Should
Embodiment by temperature inspection determine liquid level rise reason whether be due to reactor RHR system pilot operated safety valve
Phenomenon is leaked in occurring and is caused.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：It is cold to reactor
But the packing of the shower valve of agent system performs loose joint inspection, is revealed if there is packing, it is determined that pressure release case liquid level rise reason
For the packing leakage of the shower valve of reactor coolant loop；Loose joint is performed to the packing of the isolating valve of reactor coolant loop
Check, revealed if there is packing, it is determined that pressure release case liquid level rise reason is the disk of the isolating valve of reactor coolant loop
Root is revealed.The embodiment to packing loose joint inspection by determining whether liquid level rise reason is due to reactor coolant loop
The packing of shower valve and/or isolating valve is leaked and caused.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：Judge reactor
Whether boron and water make-up system the pressure release case liquid level during opening and closing go up, and go up if there is liquid level, it is determined that pressure release
There is interior leakage phenomenon for the spray isolating valve of pressure release case in case liquid level rise reason.The embodiment is by opening and closing make-up system
Operation determine liquid level rise reason whether be due to pressure release case spray isolating valve occur in leakage phenomenon and cause.
It is described to check and determine that pressure release case liquid level rise reason includes in a kind of embodiment of the application：Perform isolation cold
But the interim operating instruction of water system cooling coil, contrasted during interim operating instruction is implemented pressure release case liquid level rise speed with
Rise speed before implementation, if liquid level rise speed changes before and after implementing, it is determined that pressure release case liquid level rise reason is
Cooling water system cooling coil is leaked.Whether operation of the embodiment by performing interim operating instruction determines liquid level rise reason
It is due to that cooling water system cooling coil is leaked and caused.
Further, the pressure release case liquid level rise speed and the determination of the corresponding relation of primary Ioops slip include：Meter
The corresponding per unit slip of per unit liquid level is calculated, according to result of calculation and historical defect data, pressure release case liquid level is determined
The corresponding relation of rise speed and primary Ioops slip, the corresponding relation includes：If pressure release case liquid level rise speed is small
In first rate designated value, then its corresponding primary Ioops slip is less than the first leakage designated value, if in pressure release case liquid level
The speed that rises is more than or equal to first rate designated value and is less than the second speed designated value, then its corresponding primary Ioops slip is
More than or equal to the first leakage designated value, if pressure release case liquid level rise speed is more than or equal to the second speed designated value,
Its corresponding primary Ioops slip is more than or equal to the second leakage designated value, wherein the first rate designated value and the second speed
Rate designated value is set according to historical defect data, and the first leakage designated value is according to the scaling step and with reference to described the
One speed designated value calculates obtained numerical value, and the second leakage designated value is according to the scaling step and with reference to described second
Speed designated value calculates obtained numerical value；According to the reading for the liquid level gauge being connected with the pressure release case, calculate and current pressure release
The corresponding slip of case liquid level rise speed, the slip is the unit identical numerical value with primary Ioops slip.
Further, the corresponding primary Ioops slip processing prediction scheme of the first leakage designated value includes：According to normal prison
The frequency monitoring pressure release case liquid level rise situation of nuclear power station is surveyed, or monitors pressure release case liquid level rise situation and tracks pressure release case liquid level
The trend of rise；Described second, which leaks the corresponding primary Ioops slip processing prediction scheme of designated value, includes：Determine in pressure release case liquid level
Existing exception is swelled, checks and determines pressure release case liquid level rise reason.The embodiment is let out by historical defect data and primary Ioops
The related running technology specification of leak rate, can learn the liquid level rise speed of one loop of nuclear power station voltage-stablizer pressure release case in what value
Belong to normal monitoring range, therefore the correlation for the corresponding primary Ioops slip of liquid level rise speed that can be read according to liquid level gauge
Running technology specification, come determine whether to monitor as usual or strengthen monitoring, and judge pressure release case liquid level rise whether occur it is different
Often.
The beneficial effect of the application is：By the way that voltage-stablizer pressure release case liquid level rise speed is converted to one time of existing standard
Road slip, solves the voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station and goes up without typical problem, by setting up
Liquid level rise speed and primary Ioops slip handle the matching relationship of prediction scheme so that can be obtained after liquid level rise speed is obtained
Know corresponding processing prediction scheme, solve the voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station and go up without intervention means
Problem, so as to reduce unnecessary valve Disintegration overhaul, and then reduces radioactive dose and the radiation of nuclear power station staff
Property waste.
Brief description of the drawings
Fig. 1 is a kind of one loop of nuclear power station pressurizer system information processing method schematic flow sheet of embodiment of the application；
Fig. 2 is primary Ioops voltage-stablizer pressure release case and its annexation schematic diagram；
Fig. 3 lets out for a kind of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed of embodiment of the application with primary Ioops
Leak rate and its corresponding relation schematic diagram for handling prediction scheme；
Fig. 4 is a kind of too fast investigation of the one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed of embodiment of the application
The schematic flow sheet of processing method；
Fig. 5 is too fast rear basic for a kind of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed of embodiment of the application
Reason investigation processing schematic diagram；
Fig. 6 to Fig. 8 is voltage-stablizer pressure release case calculating scale diagrams in a kind of embodiment of the application, and wherein Fig. 6 is pressure release
The appearance schematic diagram of case, Fig. 7 for the profile of pressure release case side schematic view, Fig. 8 for pressure release case profile from the section in A-A directions
Schematic diagram.
Embodiment
The mentality of designing itself asked is that voltage-stablizer pressure release case liquid level rise speed is scaled into primary Ioops slip, formulates and closes
The pressure release case liquid level rise limit value of reason, at the same workflow based on pressure release case analyse in depth there is provided it is a kind of comprehensively, it is specific, can
The too fast reason investigation of voltage-stablizer pressure release case liquid level rise speed and processing method of operation so that when liquid level goes up and exceedes limit value
It can perform related prediction scheme.
The present invention is described in further detail below by embodiment combination accompanying drawing.
Embodiment 1：
As shown in figure 1, the present embodiment provides a kind of one loop of nuclear power station pressurizer system information especially pressure release case liquid level
The processing method of information, comprises the following steps S101~S103：
To should determine that step S101, according to the structural relation of pressure release case liquid level during normal operation and pressure release case,
Determine the corresponding relation of pressure release case liquid level rise speed and primary Ioops slip；
Obtaining step S102, obtains current pressure release case liquid level rise speed；
Scaling step S103, current pressure release case liquid level rise speed is converted to leaked with primary Ioops according to the corresponding relation
The unit identical numerical value of rate；
Matching step S104, according to the alignment processing prediction scheme of primary Ioops slip, set up pressure release case liquid level rise speed with
Primary Ioops slip handles the matching relationship of prediction scheme.
Specifically, for voltage-stablizer pressure release case liquid level rise speed is converted into primary Ioops slip, it is necessary to first understand one time
The function and structure feature of road slip and voltage-stablizer pressure release case.
One loop of nuclear power station slip be defined as in the unit interval by primary Ioops system pressure boundary leaking to secondary circuit or
The cooling agent total amount F of other systemsp, it can be divided into quantitative slip again（Confirmable leakage）FqWith non-quantitation slip（Can not
The leakage of determination）FnqTwo parts.Wherein, quantitative slip FqRefer to the leakage of certain determination position having contemplated that in the design, it
Be collected into the container specified, and flow can be determined.Non-quantitation leaks FnqRefer to that above-mentioned definition does not include other
In the case of leakage, this is that a kind of position is uncertain or position is determined but the immeasurablel leakage of leakage flow.
The need for radioactive shield, nuclear power station has strict limit value to primary Ioops slip.Obviously, to non-quantitation
The limit value of leakage will be provided more strictly, non-quantitation slip F if certain nuclear power station unit station running technology specificationsnqMore than 230L/
H or total slip FpDuring more than 2300L/h, it is both needed at the appointed time import unit and accordingly moves back anti-state.For convenience,
Only F need to be carried out in nuclear power station day-to-day operation technical specificationnqCalculate, i.e. primary Ioops and its total slip of border non-quantitation are necessarily less than
230L/h。
For voltage-stablizer pressure release case, under Operation at full power operating mode, pressure release case can receive the 110% voltage-stablizer vapor space
Steam, i.e., opened in pressurizer safety valve in time of 30 seconds, pressure release case can about receive 1.7 tons of steam, in the case
Pressure is no more than 45bar.a in pressure release case pressure release case, and temperature is no more than 93 DEG C.But the limited volume of pressure release case, it can not connect
The steam successively discharged by pressurizer safety valve.
As shown in Fig. 2 common pressure release case is a Horizontal cylinder shape low pressure vessel, its left and right ends is sealed for hemispherical
Head, total measurement (volume) about 37m3, under normal circumstances, water level is the 65% of total height in case, and water temperature maintains 40 DEG C.Top is filled with nitrogen
Gas, rated pressure is 1.2bar.a（Absolute pressure value）.The purpose of inflated with nitrogen is the hydrogen for preventing from containing in the steam discharged by voltage-stablizer
Gas produces detonation gas with the oxygen mix in air.Hydrogen and oxygen concentration that periodically sampling analysis is assembled out of case, and will
It is discharged into nuclear island draining and draining system（PRE）.
Pressure release case is built with one by reactor boron and water make-up system（REA）The shower of water supply, a piece-root grafting RPE systems
Hydrophobic pipeline, the former is used for cooling down pressure release case when safety valve discharge, and the latter is for the draining when cistern water level is high.The water of case
There is one in space by component cooling water system（RRI）The cooling coil of water supply, one is being equipped with close to bottom in the axial direction
Bubbling pipe, this root pipe is connected with voltage-stablizer blowdown line.Pressure release upper box part presses case superpressure provided with two rupture disks to prevent stopping leak, quick-fried
The emission of broken disk enters in the air of containment, and its relieving capacity is equal to three safety valve discharge capacity sums of voltage-stablizer.
During normal operation, the RRI system equipments cooling water that the water in pressure release case flows through in coiled pipe in case is incessantly
Cooling.If water temperature is more than 60 DEG C, alarm signal is sent, operator wants manually opened voltage-stablizer pressure release case spray isolating valve,
Pressure release case is cooled down through shower penetrating from the desalination degassed water of REA systems, maximum spray flux is 13.6m3/h.If water
Position is too high, then opens the drain valve in pressure release bottom portion to RPE System drainages, but water temperature height to 65 DEG C when, forbid pressure release case automatically
The drain valve of bottom is opened, to avoid high-temperature water from flowing to RPE systems.
By the structural analysis to voltage-stablizer pressure release case, pressure release case liquid level is computed in 2.0~2.2m during normal operation
Spherical crown understands that every millimeter of liquid level correspondence slip is 14.414 to 15.678 liters with columnar volume, and due to nuclear plant safety
It is required that it is very high, therefore the calculating does not consider that cooling coil volume influences, it is too conservative.Here what is be specifically related to is calculated as follows：
As shown in Figure 6 to 8, voltage-stablizer pressure release case is made up of two Side Volume identical spherical crowns and mediate cylindrical tank body,
Wherein, spherical crown volume in left side is set to V2, right side spherical crown volume is set to V3, mediate cylindrical volume is set to V1, using left side spherical crown as
Example, if the radius of spherical crown is r, the line segment and horizontal direction shape that intersection point to the centre of sphere between spherical crown and cylinder is formed are into θ angle, ball
The maximum distance of hat and cylinder is the intersection point between a, intersection point and spherical crown and cylinder of the maximum distance correspondence on spherical crown
The distance between be ri, it is assumed that the height L of cylinder, a diameter of D of round sides, while it is also supposed that on A-A sections cylinder
The area of body is S1, left side spherical corona's area is S2, right side spherical corona's area is S3, the width of cylinder is D in the sectioni.Normal fortune
Pressure release case liquid level is in 2.0~2.2m, when calculating jar liquid level rise speed, the change of the sectional area gone up per 1mm between the departure date
It is negligible.Therefore liquid surface area when liquid level is 2.0m and 2.2m is calculated respectively, you can draw liquid level often rise 1mm bodies
Long-pending rate of change, so as to calculate liquid level rise speed.
When pressure release case liquid level is h, calculating process is as follows：
ByIt can draw
ByIt can draw
Because both sides spherical crown volume is identical, it can draw
When liquid level is h, that is, have
It was found from Fig. 6-8, L=60+1220+1220+1960+60=4520mm=4.52m, D=3000mm=3m, a=856-60=
796mm=0.796m。
1）As h=2.0m, data are substituted into formula（3）、（5）、（6）, then S is calculated1=15.678m2, i.e. now liquid level
Often go up 1mm, volume increase V1=15.678L；
2）As h=2.2m, data are substituted into formula 3,5,6, then calculate S2=14.414m2, i.e., now liquid level often goes up
1mm, volume increase V2=14.414L。
Accordingly, voltage-stablizer pressure release case current level rise speed can be scaled primary Ioops slip, i.e. according to letting out
The reading of the connected liquid level gauge of case is pressed, learns rise speed for K millimeters daily（mm/d）, then corresponding primary Ioops slip
FcurrentFor equation below：
Fcurrent=K×（14.414~15.678）/ 24, unit：Liter per hour（L/h）（7）
It can be identified as so as to the corresponding relation of pressure release case liquid level rise speed and primary Ioops slip：
If pressure release case liquid level rise speed is less than first rate designated value, its corresponding primary Ioops slip is small
Designated value is leaked in first；
If pressure release case liquid level rise speed is more than or equal to first rate designated value and less than the second speed designated value,
Then its corresponding primary Ioops slip is more than or equal to the first leakage designated value；
If pressure release case liquid level rise speed is more than or equal to the second speed designated value, its corresponding primary Ioops leakage
Rate is more than or equal to the second leakage designated value；
Wherein, first rate designated value and the second speed designated value are set according to historical defect data, historical defect data
Source can be, data for example collected in nuclear power station overhaul, or the data collected during nuclear power station day-to-day operation.First
It is according to formula to leak designated value（7）And the numerical value that the calculating of first rate designated value is obtained is combined, the second leakage designated value is root
According to formula（7）And combine the numerical value that the calculating of the second speed designated value is obtained.
In practice, according to the historical defect data of record, multiple one loop of nuclear power station voltage-stablizer pressure release case liquid levels go up fast
Rate repeatedly reaches 10mm/d or so, primary Ioops slip is scaled for 6.25L/h, apart from primary Ioops slip standard（It is i.e. foregoing
230L/h）Farther out.Moreover, after the increase of pressure release case liquid level rise speed, voltage-stablizer endless tube hydrogen content will be caused to rise, and according to
Historical experience, liquid level rise speed is in 10mm/d or so, and the annual voltage-stablizer endless tube nitrogen purging that performs is not more than 1 time, belongs to
In tolerance interval；In addition, nuclear power station alarm card request primary Ioops voltage-stablizer pressure release case liquid level reaches 2.42 meters of heel row of alarming value
Water is to 2 meters or so, and 0.42 meter of drain height is calculated, draining frequency is 42 days, falls within acceptable model if pressing slip 10mm/d
In enclosing.And primary Ioops voltage-stablizer pressure release case liquid level rise speed reaches the feelings more than 10mm/d and less than 15mm/d in historical experience
Condition is less, need to start to strengthen paying close attention to and tracking trend；Learnt according to historical experience, multiple one loop of nuclear power station voltage-stablizer pressure release casees
Liquid level rise speed is about 15mm/d to the maximum.Therefore, first rate designated value can be set to 10mm/d, its corresponding first leakage
Designated value is 6.25L/h, and the second speed designated value is 15mm/d, and the second leakage designated value is 9.375L/h.
With reference to nuclear power station running technology specification, it can be deduced that primary Ioops voltage-stablizer pressure release case liquid level rise speed control mark
Relation between standard, i.e. rise speed-primary Ioops slip-processing prediction scheme, specifically, primary Ioops slip be less than
During 6.25L/h, according to the frequency monitoring pressure release case liquid level rise situation of normal monitoring nuclear power station；Primary Ioops slip be more than etc.
In 6.25L/h and less than 9.375L/h when, strengthen monitoring pressure release case liquid level rise situation simultaneously track pressure release case liquid level rise become
Gesture；When primary Ioops slip is more than or equal to 9.375L/h, it is determined that pressure release case liquid level, which goes up, exception occurs, checks and determines to let out
Case liquid level rise reason is pressed, related prediction scheme can be started and handled.For understanding directly perceived, the form shown in Fig. 3 is referred to.
The present embodiment is by the way that voltage-stablizer pressure release case liquid level rise speed to be converted to the primary Ioops slip of existing standard, solution
The voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station of having determined goes up without typical problem, and the liquid level for passing through foundation goes up speed
Rate handles the matching relationship of prediction scheme with primary Ioops slip so that can learn corresponding processing after liquid level rise speed is obtained
Prediction scheme, so that operator can handle pre- to obtained one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed using corresponding
Case.
Embodiment 2：
As shown in figure 4, a kind of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed of the present embodiment offer is too fast
Processing method is investigated, including：
Establishment step S401, according to the corresponding relation of pressure release case liquid level rise speed and primary Ioops slip, will currently let out
Pressure case liquid level rise speed is converted to the unit identical numerical value with primary Ioops slip, and sets up pressure release case liquid level rise speed
The matching relationship of prediction scheme is handled with primary Ioops slip；
Step S403 is selected, according to numerical value, corresponding primary Ioops slip processing prediction scheme is selected in the matching relationship；
Step S405 is investigated, the primary Ioops slip processing prediction scheme selected is performed, and it is determined that pressure release case liquid level goes up
When occurring abnormal, check and determine pressure release case liquid level rise reason.
Wherein step S401's and S403 implements the associated description that process refers to embodiment 1, no longer repeats.Below
Specifically describe step S405.
As shown in Figure 2, the water source collected by voltage-stablizer pressure release case has four, is respectively：
1. RCP/RCV/RRA systems safety valve water；
2. RCP valve packings draw leakage water；
3. REA shower waters；
4. RRI cooling coils water.
According to each road water, progress coherence check can by taking No. 1 unit L1 of certain nuclear power station as an example, and with specialized department
Know：
For 1. RCP/RCV/RRA systems safety valve water, it includes RCP system pilot operated safety valves（Such as RCP020/
021/022VP）, RCV system pilot operated safety valves（Such as RCV201VP）, RRA system pilot operated safety valves（Such as RRA018/
115VP）With RCV system spring loaded safety valves（Such as RCV252VP）Water leakage.It is widely used at present in the important system of nuclear power station
The SEBIM safety valves that pilot operated safety valve in system equipment provides for SEBIM companies of France, it is for pressure vessel and pressure-bearing pipe
Road provides overpressure protection, has the advantages that common safety valve is incomparable, its special pilot control design not only seal it is tight,
Adjust precision height, be swift in motion, and solve the significant trouble hidden danger for not returning seat after conventional security valve is opened well, be core
The important leverage of group of motors safe and stable operation.
Accordingly, for 1. RCP/RCV/RRA systems safety valve water, it can be determined by checking as follows in pressure release case liquid level
Rise reason：
I. safety valve jumping up pressure fixed value checking and air seal test are performed, if verification and result of the test are unqualified, really
Determine the spring loaded safety valve that pressure release case liquid level rise reason is RCV systems and interior leakage phenomenon occur, this investigation action is generally big
Used when repairing；
II. the pilot operated safety valve discharge pipe line temperature of RCV systems is checked, judges whether the trend of its temperature goes up
Phenomenon, if going up, it is determined that pressure release case liquid level rise reason interior leakage phenomenon occurs for the pilot operated safety valve of RCV systems,
This investigation action is generally used in current check；
III. check RRA systems pilot operated safety valve discharge pipe line temperature, judge its temperature trend whether occur on
Rise phenomenon, if going up, it is determined that it is existing that pressure release case liquid level rise reason interior leakage occurs for the pilot operated safety valve of RRA systems
As this investigation action is generally used in current check；
IV. the pilot operated safety valve discharge pipe line temperature of RCP systems is checked, judges whether the trend of its temperature goes up
Phenomenon, if going up, it is determined that pressure release case liquid level rise reason interior leakage phenomenon occurs for the pilot operated safety valve of RCP systems,
This investigation action is generally used in current check.
Draw leakage water for 2. RCP valve packings, it includes RCP stabilizer shower valves（Such as RCP001/002VP）With every
From valve（Such as RCP102/103/202/203/302/303VP）Packing draw leak.
Draw leakage water accordingly, for 2. RCP valve packings, can determine that pressure release case liquid level goes up by checking as follows former
Cause：
I. loose joint inspection is performed to the packing of the shower valves of RCP systems, that is, opens packing and draw leakage loose joint and check whether there is disk
Root leakage vestige, boron crystallization etc., if it is present determining packing of the pressure release case liquid level rise reason for the shower valve of RCP systems
Leakage, this investigation action is generally used in overhaul;
II. loose joint inspection is performed to the packing of the isolating valves of RCP systems, that is, opens packing and draw leakage loose joint and check whether there is disk
Root leakage vestige, boron crystallization etc., if it is present determining packing of the pressure release case liquid level rise reason for the isolating valve of RCP systems
Leakage, this investigation action is generally used in overhaul.
For 3. REA shower waters, it is leaked to spray isolating valve from voltage-stablizer（Such as REA001/002PO）Supply water.
Accordingly, for 3. REA shower waters, pressure release case liquid level rise reason can be determined by checking as follows：
Judge whether REA systems pressure release case liquid level during opening and closing goes up, go up if there is liquid level, it is determined that
There is interior leakage phenomenon for the spray isolating valve of pressure release case in pressure release case liquid level rise reason, and this investigation action is generally in current check
Shi Caiyong.
For 4. RRI cooling coils water, it is because corrosive pipeline perforation leakage causes water.
Accordingly, for 4. RRI cooling coils water, pressure release case liquid level rise reason can be determined by checking as follows：
Perform the interim operating instruction of RRI system cooling coils（TOI）, pressure release is contrasted during interim operating instruction is implemented
Case liquid level rise speed and the rise speed before implementation, if liquid level rise speed changes before and after implementing, it is determined that pressure release
Case liquid level rise reason is leaked for RRI systems cooling coil, and this investigation action is generally used in current check.
The exceeded basic original of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed can be determined from above-mentioned analysis
Processing scheme after cause, the action of reason investigation and investigation, i.e. the very fast question processing method of liquid level rise speed can be directed to foregoing
Four tunnel waters separately verify and check, as shown in Figure 5.
The present embodiment is by the way that voltage-stablizer pressure release case liquid level rise speed to be converted to the primary Ioops slip of existing standard, solution
The voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station of having determined goes up without typical problem, and the liquid level for passing through foundation goes up speed
Rate handles the matching relationship of prediction scheme with primary Ioops slip so that can learn corresponding processing after liquid level rise speed is obtained
Prediction scheme, and can be checked when rise speed is too fast and determine liquid level rise reason, so as to according to the reason for investigating out using pair
The processing method answered solves problem, it is seen that the present embodiment solves the voltage-stablizer pressure release case liquid level existed for a long time in current nuclear power station
The problem of going up without intervention means, so as to reduce unnecessary valve Disintegration overhaul, and then reduce nuclear power station staff
Radioactive dose and radwaste.
To sum up embodiment, it is known that itself please pass through 1）Relation between pressure release case water level rise speed and primary Ioops slip
Quantitative analysis and 2）Pressure release case water level goes up may source analysis, i.e. RCP/RCV/RRA systems safety valve water, RCP valve disks
Root draws leakage water, REA shower waters, RRI cooling coil waters, has formulated pressure release case water level rise speed control standard and the speed that goes up
Rate exceeded rear basic reason investigation and processing prediction scheme, i.e.,：Rise speed<During 10mm/d, monitored according to normal frequency；Rise speed
>=10mm/d and≤15mm/d, strengthen monitoring and tracking trend；Rise speed >=15mm/d, start-up check prediction scheme.So as to solve
The significant problem that the voltage-stablizer pressure release case water level existed for a long time in nuclear power station goes up without standard, without intervention means, is reduced not
Necessary valve Disintegration overhaul, and the radioactive dose and radwaste of staff is reduced, with important economy and society
Can benefit.
Above content is to combine the further description that specific embodiment is made to the application, it is impossible to assert this Shen
Specific implementation please is confined to these explanations.For the application person of an ordinary skill in the technical field, do not taking off
On the premise of from the application design, some simple deduction or replace can also be made.
Claims (13)
1. a kind of one loop of nuclear power station pressurizer system information processing method, it is characterised in that including：
To should determine that step, according to the structural relation of pressure release case liquid level during normal operation and pressure release case, pressure release is determined
The corresponding relation of case liquid level rise speed and primary Ioops slip；
Obtaining step, obtains current pressure release case liquid level rise speed；
Scaling step, current pressure release case liquid level rise speed is converted to the list with primary Ioops slip according to the corresponding relation
Position identical numerical value；
Match establishment step, according to the alignment processing prediction scheme of the primary Ioops slip, set up pressure release case liquid level rise speed with
Primary Ioops slip handles the matching relationship of prediction scheme；Described pair should determine that step includes：Calculate per unit liquid level corresponding every
Unit slip, according to result of calculation and historical defect data, determines pressure release case liquid level rise speed and primary Ioops slip
Corresponding relation；
The obtaining step includes：According to the reading for the liquid level gauge being connected with the pressure release case, obtain in current pressure release case liquid level
Rise speed；
The scaling step includes：Slip corresponding with current pressure release case liquid level rise speed is calculated, the slip is
With the unit identical numerical value of primary Ioops slip；
The pressure release case liquid level rise speed and the determination step of the corresponding relation of primary Ioops slip include：
If pressure release case liquid level rise speed is less than first rate designated value, it is determined that its corresponding primary Ioops slip is small
Designated value is leaked in first；
If pressure release case liquid level rise speed is more than or equal to first rate designated value and is less than the second speed designated value, really
Its fixed corresponding primary Ioops slip is more than or equal to the first leakage designated value；
If pressure release case liquid level rise speed is more than or equal to the second speed designated value, it is determined that its corresponding primary Ioops leakage
Rate is more than or equal to the second leakage designated value；
The first rate designated value and the second speed designated value are set according to historical defect data, the first leakage designated value
To calculate obtained numerical value according to the scaling step and with reference to the first rate designated value, the second leakage designated value is
Calculate according to the scaling step and with reference to the second speed designated value obtained numerical value.
2. one loop of nuclear power station pressurizer system information processing method as claimed in claim 1, it is characterised in that the pressure release
The centre that is shaped as of case is that cylindrical and both sides are volume identical spherical crown, and the liquid level during normal operation is 2 meters
To 2.2 meters, the corresponding per unit slip of per unit liquid level that calculates includes：When calculating liquid level is 2 meters and 2.2 meters respectively
Liquid surface volume, further according to spherical crown volume and cylinder volume formula, calculate liquid level often go up 1 millimeter when volume change
Rate, so as to obtain corresponding primary Ioops slip.
3. one loop of nuclear power station pressurizer system information processing method as claimed in claim 1, it is characterised in that described first
Speed designated value is 10 millimeters daily, and the first leakage designated value is 6.25 liter per hour, and the second speed designated value is
15 millimeters daily, and the second leakage designated value is 9.375 liter per hour.
4. one loop of nuclear power station pressurizer system information processing method as claimed in claim 1, it is characterised in that
Described first, which leaks the corresponding primary Ioops slip processing prediction scheme of designated value, includes：According to the frequency of normal monitoring nuclear power station
Monitor pressure release case liquid level rise situation, or monitoring pressure release case liquid level rise situation and the trend for tracking the rise of pressure release case liquid level；
Described second, which leaks the corresponding primary Ioops slip processing prediction scheme of designated value, includes：Determine that the rise of pressure release case liquid level occurs different
Often, check and determine pressure release case liquid level rise reason.
5. a kind of too fast investigation processing method of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed, it is characterised in that
Including：
Establishment step, according to the corresponding relation of pressure release case liquid level rise speed and primary Ioops slip, by current pressure release case liquid level
Rise speed is converted to the unit identical numerical value with primary Ioops slip, and sets up pressure release case liquid level rise speed and primary Ioops
Slip handles the matching relationship of prediction scheme；
Step is selected, according to numerical value, corresponding primary Ioops slip processing prediction scheme is selected in the matching relationship；
Step is investigated, the primary Ioops slip processing prediction scheme selected is performed, and it is abnormal in the appearance it is determined that pressure release case liquid level goes up
When, check and determine pressure release case liquid level rise reason；
The pressure release case liquid level rise speed and the determination of the corresponding relation of primary Ioops slip include：
The corresponding per unit slip of per unit liquid level is calculated, according to result of calculation and historical defect data, pressure release is determined
The corresponding relation of case liquid level rise speed and primary Ioops slip, the corresponding relation includes：The speed if pressure release case liquid level goes up
Rate is that then its corresponding primary Ioops slip is less than the first leakage designated value, if pressure release case less than first rate designated value
Liquid level rise speed is more than or equal to first rate designated value and is less than the second speed designated value, then its corresponding primary Ioops is let out
Leak rate is more than or equal to the first leakage designated value, if pressure release case liquid level rise speed is to be specified more than or equal to the second speed
Value, then its corresponding primary Ioops slip be more than or equal to second leakage designated value, wherein the first rate designated value and
Second speed designated value is set according to historical defect data, and the first leakage designated value is will be current according to the corresponding relation
Pressure release case liquid level rise speed is converted to the unit identical numerical value with primary Ioops slip and specified with reference to the first rate
Value calculates obtained numerical value, and the second leakage designated value is by current pressure release case liquid level rise speed according to the corresponding relation
Be converted to the unit identical numerical value with primary Ioops slip and calculate obtained numerical value with reference to the second speed designated value；
According to the reading for the liquid level gauge being connected with the pressure release case, let out corresponding with current pressure release case liquid level rise speed is calculated
Leak rate, the slip is the unit identical numerical value with primary Ioops slip.
6. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
Safety valve jumping up pressure fixed value checking and air seal test are performed, if verification and result of the test are unqualified, it is determined that pressure release
Case liquid level rise reason is that the spring loaded safety valve of chemical volume control system interior leakage phenomenon occurs.
7. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
Check the pilot operated safety valve discharge pipe line temperature of chemical volume control system, judge its temperature trend whether occur on
Rise phenomenon, if going up, it is determined that pressure release case liquid level rise reason is that the pilot operated safety valve of chemical volume control system goes out
Leakage phenomenon in existing.
8. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
The pilot operated safety valve discharge pipe line temperature of reactor RHR system is checked, judges whether the trend of its temperature occurs
Rise phenomenon, if going up, it is determined that pressure release case liquid level rise reason is safe for the pilot-operated type of reactor RHR system
Valve leaks phenomenon in occurring.
9. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
Check reactor coolant loop pilot operated safety valve discharge pipe line temperature, judge its temperature trend whether occur on
Rise phenomenon, if going up, it is determined that pressure release case liquid level rise reason goes out for the pilot operated safety valve of reactor coolant loop
Leakage phenomenon in existing.
10. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
Loose joint inspection is performed to the packing of the shower valve of reactor coolant loop, revealed if there is packing, it is determined that pressure release
Case liquid level rise reason is revealed for the packing of the shower valve of reactor coolant loop；
Loose joint inspection is performed to the packing of the isolating valve of reactor coolant loop, revealed if there is packing, it is determined that pressure release
Case liquid level rise reason is revealed for the packing of the isolating valve of reactor coolant loop.
11. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
Judge whether reactor boron and water make-up system the pressure release case liquid level during opening and closing go up, if there is in liquid level
Rise, it is determined that pressure release case liquid level rise reason interior leakage phenomenon occurs for the spray isolating valve of pressure release case.
12. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that in the investigation step, it is described to check and determine that pressure release case liquid level rise reason includes：
The interim operating instruction of xegregating unit cooling water system cooling coil is performed, contrasts and lets out during interim operating instruction is implemented
Rise speed before pressing case liquid level rise speed and implementing, if liquid level rise speed changes before and after implementing, it is determined that let out
Case liquid level rise reason is pressed to be leaked for component cooling water system cooling coil.
13. the too fast investigation processing side of one loop of nuclear power station voltage-stablizer pressure release case liquid level rise speed as claimed in claim 5
Method, it is characterised in that
Described first, which leaks the corresponding primary Ioops slip processing prediction scheme of designated value, includes：According to the frequency of normal monitoring nuclear power station
Monitor pressure release case liquid level rise situation, or monitoring pressure release case liquid level rise situation and the trend for tracking the rise of pressure release case liquid level；
Described second, which leaks the corresponding primary Ioops slip processing prediction scheme of designated value, includes：Determine that the rise of pressure release case liquid level occurs different
Often, check and determine pressure release case liquid level rise reason.
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| CN107369480B (en) * | 2016-05-12 | 2019-09-17 | 福建宁德核电有限公司 | A kind of measurement method and device of nuclear power station loop leakage rate |
| CN109018287B (en) * | 2018-08-01 | 2020-03-27 | 中国船舶科学研究中心（中国船舶重工集团公司第七0二研究所） | Pressure stabilizer pressure relief system for deep-sea nuclear power underwater platform |
| CN112464134B (en) * | 2019-09-09 | 2023-09-08 | 中核核电运行管理有限公司 | Nuclear power plant primary loop leakage rate sub-working condition calculation method |
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| CN110828007B (en) * | 2019-11-18 | 2021-06-29 | 中国核动力研究设计院 | Special voltage stabilizer and pressure control system for reactor irradiation examination loop |
| CN111524623B (en) * | 2020-04-30 | 2022-02-22 | 中国核动力研究设计院 | Constant value and arrangement method for safety valve of voltage stabilizer |
| CN113571211B (en) * | 2021-07-06 | 2023-12-19 | 中国核电工程有限公司 | Nuclear power system and method and primary loop system thereof as well as reactor overpressure protection system and method |
| CN114038592B (en) * | 2021-10-12 | 2024-03-15 | 中广核陆丰核电有限公司 | Nuclear power plant primary loop leakage rate monitoring method and device |
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Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2915012B2 (en) * | 1989-08-30 | 1999-07-05 | 株式会社日立製作所 | Nuclear power plant |
| CN101706039A (en) * | 2009-11-24 | 2010-05-12 | 中国核动力研究设计院 | Method for monitoring pressure pipeline leakage acoustic emission in nuclear power plant and monitoring system thereof |
| CN102543232A (en) * | 2011-10-24 | 2012-07-04 | 上海电力学院 | Combined method for controlling water level and pressure of voltage stabilizer for nuclear power plant of pressurized water reactor |
| CN203070790U (en) * | 2013-01-14 | 2013-07-17 | 上海核工程研究设计院 | System for quantitatively measuring coolant leakage rate of primary loop of pressurized water reactor nuclear power plant |
| CN103726834A (en) * | 2012-10-10 | 2014-04-16 | 北京格瑞迪斯石油技术有限公司 | Sustained casing pressure diagnosis device and method |
-
2013
- 2013-09-11 CN CN201310413734.8A patent/CN104425045B/en active Active
Patent Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2915012B2 (en) * | 1989-08-30 | 1999-07-05 | 株式会社日立製作所 | Nuclear power plant |
| CN101706039A (en) * | 2009-11-24 | 2010-05-12 | 中国核动力研究设计院 | Method for monitoring pressure pipeline leakage acoustic emission in nuclear power plant and monitoring system thereof |
| CN102543232A (en) * | 2011-10-24 | 2012-07-04 | 上海电力学院 | Combined method for controlling water level and pressure of voltage stabilizer for nuclear power plant of pressurized water reactor |
| CN103726834A (en) * | 2012-10-10 | 2014-04-16 | 北京格瑞迪斯石油技术有限公司 | Sustained casing pressure diagnosis device and method |
| CN203070790U (en) * | 2013-01-14 | 2013-07-17 | 上海核工程研究设计院 | System for quantitatively measuring coolant leakage rate of primary loop of pressurized water reactor nuclear power plant |
Non-Patent Citations (1)
| Title |
|---|
| 压水堆核电厂一回路冷却剂泄漏率计算的优化;詹勇杰，何子帅，潘泽飞;《核动力工程》;20090430;第30卷(第2期);86页第3-4段 * |
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