CN110689973B - Nuclear power station primary loop depressurization control method under heat transfer pipe rupture accident - Google Patents
Nuclear power station primary loop depressurization control method under heat transfer pipe rupture accident Download PDFInfo
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- G21C9/00—Emergency protection arrangements structurally associated with the reactor, e.g. safety valves provided with pressure equalisation devices
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
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- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
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Abstract
The invention relates to a primary loop depressurization control method of a nuclear power station under a heat transfer pipe fracture accident, which comprises the following steps: s1, acquiring fault information of a nuclear power station to calculate an enabling signal; s2, judging whether the average temperature of the coolant meets the supercooling degree requirement or not, and obtaining a cooling judgment result; s3, acquiring isolation information of the steam generator, and calculating to obtain an update signal by combining an enabling signal and a cooling judgment result; s4, obtaining a pressure reference value of the pressure control system based on the updating signal according to the secondary side pressure of the steam generator and the pressure set value of the pressure control system; s5, obtaining the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so that the pressure of the loop is reduced. Compared with the prior art, the method and the device have the advantages that the pressure reference value of the pressure control system can be automatically obtained by calculating and processing the monitoring data of the nuclear power station, the problems of inaccurate manual operation and overlong judging time are avoided, and the accuracy and the speed of emergency treatment are improved.
Description
Technical Field
The invention relates to the technical field of heat transfer tube rupture accident treatment of a steam generator of a nuclear power station, in particular to a primary loop depressurization control method of the nuclear power station under the heat transfer tube rupture accident.
Background
The rupture of the heat transfer tubes of the steam generator in the nuclear power station means that one or more heat transfer tubes in the steam generator are ruptured, so that the integrity of the pressure boundary of a primary circuit is lost, the primary circuit is communicated with the secondary circuit, and a nuclear leakage accident is caused. For this reason, the existing emergency treatment mode is usually that after the heat transfer tube breaks down, an operator enters a standard operation procedure according to the requirements of a corresponding alarm or technical specification program to control the accident, and the main control strategies are as follows:
1. identifying and isolating a faulty steam generator;
2. under the condition of ensuring the supercooling degree, the primary circuit is controlled to reduce the pressure as soon as possible, so that leakage is reduced and eliminated;
3. when the pressures of the first loop and the second loop tend to be balanced, the reactor is withdrawn to a safe state by adopting a mode of synchronously reducing the pressure of the first loop and the second loop.
The process requires operators to continuously judge the main control state and then control accidents by adopting corresponding means, but in the actual manual operation process, the steps are complicated, so that the judging operation time is long, the manual intervention measures are not timely and accurate enough, especially when the manual depressurization of a loop is carried out, the operation difficulty is further increased due to excessive conditions required to be judged, the pressure of the loop cannot be timely, accurately and quickly reduced, and more serious nuclear leakage accidents are easily caused.
Disclosure of Invention
The invention aims to overcome the defects of the prior art and provide a primary loop depressurization control method of a nuclear power station under the accident of breaking a heat transfer pipe.
The aim of the invention can be achieved by the following technical scheme: a nuclear power station primary loop depressurization control method under a heat transfer pipe rupture accident comprises the following steps:
s1, acquiring fault information of a nuclear power station to calculate an enabling signal;
s2, acquiring the secondary side pressure of the steam generator and the average temperature of the coolant, and judging whether the average temperature of the coolant meets the supercooling degree requirement or not to obtain a cooling judgment result;
s3, acquiring isolation information of the steam generator, and calculating to obtain an update signal by combining an enabling signal and a cooling judgment result;
s4, obtaining a pressure reference value of the pressure control system based on the updating signal according to the secondary side pressure of the steam generator and the pressure set value of the pressure control system;
s5, obtaining the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so that the pressure of the loop is reduced.
Further, in the step S1, the fault information of the nuclear power plant includes a heat transfer pipe rupture signal and a shutdown signal, the enable signal is obtained by performing logical and calculation on the heat transfer pipe rupture signal and the shutdown signal, when the shutdown signal and the heat transfer pipe rupture signal are both "1", the enable signal is "1", otherwise, the enable signal is "0".
Further, the step S2 specifically includes the following steps:
s21, acquiring a secondary side pressure of the steam generator, and calculating to obtain the saturation temperature of the steam generator according to a relation function of the pressure and the saturation temperature of the steam generator;
s22, acquiring the average temperature of the coolant, and calculating the cooling degree deviation by combining the saturation temperature of the steam generator;
s23, judging whether the cooling degree deviation meets the requirement:
-δ＜TP＜+δ
wherein TP represents supercooling degree deviation, and delta represents supercooling degree fluctuation margin;
and S24, if the supercooling degree deviation TP meets the requirement in the step S23, the average temperature of the coolant reaches the supercooling degree requirement to obtain a cooling judgment result of 1, otherwise, the average temperature of the coolant does not reach the supercooling degree requirement to obtain the cooling judgment result of 0.
Further, in the step S22, the calculation formula of the cooling degree deviation TP is as follows:
TP＝Tav-(T2-Ts)
T2＝F(P2)
wherein Tav represents the average temperature of the coolant, T2 represents the saturation temperature of the failed steam generator, ts represents the standard value of the supercooling degree, P2 represents the secondary side pressure of the steam generator, F (x) is a function of the relationship between the steam generator pressure and the saturation temperature, x is the steam generator pressure, and F (P2) is a function of the relationship between the secondary side pressure of the steam generator and the saturation temperature.
Further, in the step S3, the update signal is obtained by performing logical and calculation on the steam generator isolation information, the enable signal and the cooling judgment result, when the steam generator isolation information, the enable signal and the cooling judgment result are all "1", the update signal is "1", otherwise, the update signal is "0".
Further, in the step S3, the steam generator isolation information includes a main steam isolation valve closing signal, a steam-driven auxiliary water supply pump steam supply valve closing signal, and a drain valve closing signal, where the signal value of the steam generator isolation information is obtained by performing logic and computation on the main steam isolation valve closing signal, the steam-driven auxiliary water supply pump steam supply valve closing signal, and the drain valve closing signal, and when the main steam isolation valve closing signal, the steam-driven auxiliary water supply pump steam supply valve closing signal, and the drain valve closing signal are all "1", the steam generator isolation information is "1", otherwise the steam generator isolation information is "0".
Further, the step S4 specifically includes the following steps:
s41, taking the update signal as a switching condition, executing a step S42 when the update signal is 1, otherwise executing a step S43;
s42, setting the secondary side pressure value of the steam generator as a pressure reference value of a pressure control system;
s43, setting the pressure set value of the pressure control system as the pressure reference value of the pressure control system, namely maintaining the pressure set value of the current pressure control system unchanged.
Further, the step S5 specifically includes the following steps:
s51, obtaining the pressure of the voltage stabilizer, and then performing difference calculation on the pressure of the voltage stabilizer and a pressure reference value of a pressure control system to obtain pressure deviation;
s52, controlling the opening value of the spraying valve of the voltage stabilizer through PID according to the pressure deviation, enabling the spraying valve to spray, and reducing the pressure of the primary loop.
Compared with the prior art, the invention has the following advantages:
1. the invention takes the logical AND result of the heat transfer tube rupture signal and the shutdown signal as the enabling signal, combines the evaporator isolation information and the cooling judgment result as the updating signal for switching the pressure reference value of the set pressure control system, can automatically reduce the pressure after ensuring effective cooling, and can quickly and timely and reliably meet the requirement of stopping nuclear leakage.
2. The invention can quickly and accurately set the pressure reference value of the pressure control system through automatic switching, avoids the problems of inaccurate manual operation and overlong judging time, is beneficial to the subsequent timely control of the opening value of the spray valve of the pressure control system, does not influence the normal operation of the existing pressure control system, and has strong actual operability.
Drawings
FIG. 1 is a flow chart of the method of the present invention;
FIG. 2 is a schematic diagram of a system for implementing a buck control method in an embodiment;
fig. 3 is a logic structure diagram of a buck control method in an embodiment.
The figure indicates:
Detailed Description
The invention will now be described in detail with reference to the drawings and specific examples.
As shown in fig. 1, a primary loop depressurization control method for a nuclear power plant in case of a heat transfer pipe rupture accident comprises the following steps:
s1, acquiring fault information of a nuclear power station to calculate an enabling signal;
s2, acquiring the secondary side pressure of the steam generator and the average temperature of the coolant, and judging whether the average temperature of the coolant meets the supercooling degree requirement or not to obtain a cooling judgment result;
s3, acquiring isolation information of the steam generator, and calculating to obtain an update signal by combining an enabling signal and a cooling judgment result;
s4, obtaining a pressure reference value of the pressure control system based on the updating signal according to the secondary side pressure of the steam generator and the pressure set value of the pressure control system;
s5, obtaining the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so that the pressure of the loop is reduced.
As shown in fig. 2, in this embodiment, monitoring data is read by a nuclear power plant control system, programming of a loop automatic depressurization control logic of the nuclear power plant after a heat transfer pipe rupture accident is performed on a netcotrol system platform, and mutual communication between the netcotrol system program and the nuclear power plant control system is realized by using OPC communication, specifically, a variable database corresponding to an acquired monitoring point is established in the netcotrol system by using OPC as a communication tool, data acquired from the nuclear power plant control system is set as an input variable group, data output to the nuclear power plant control system is set as an output variable group, and variables calculated inside the netcotrol system are set as an intermediate variable group. On the basis of setting the completion variable group, according to the logic relation of the automatic depressurization control method, a script program is edited in the NETCONTROL system, and finally the automatic depressurization control in the emergency response after the heat transfer pipe of the steam generator breaks down.
In the embodiment, the logic structure of the buck control method is shown in fig. 3, and the specific application process includes:
(1) The method comprises the steps of carrying out real-time online monitoring on a fault diagnosis system and a nuclear power station instrument control system to obtain the state and monitoring data of a nuclear power station monitoring point:
the method comprises the steps of respectively obtaining a shutdown signal S, a voltage stabilizer pressure P1, a steam generator secondary side pressure P2, a coolant average temperature Tav, a main steam isolation valve state D1, an auxiliary feed water pump air supply valve state D2 and a sewage pipeline valve state D3 through real-time online monitoring; and acquiring a steam generator heat transfer pipeline rupture signal U through online detection of a fault diagnosis system.
(2) An automatic depressurization control method based on emergency treatment after a steam generator heat transfer tube rupture accident is realized in a NETCONTROL system. The shutdown signal S and the heat transfer tube rupture signal U trigger the coolant system to automatically reduce the average temperature of the coolant, when the heat transfer tube of the steam generator is ruptured, the shutdown signal S sent by the protection system and the heat transfer tube rupture signal U of the steam generator sent by the fault diagnosis system are obtained, and an enabling signal Q is obtained, when the heat transfer tube of the steam generator is ruptured, U is 1, otherwise U is 0; when the reactor is shut down, S is 1, otherwise S is 0; when the steam generator heat transfer tube rupture signal U and the shutdown signal S are simultaneously 1, Q is 1;
when the temperature of the coolant system is reduced to meet the supercooling degree requirement, automatically triggering the depressurization operation, and when an enabling signal Q is 1, triggering the coolant system to automatically reduce the average coolant temperature Tav to enable the coolant system temperature Tav to be reduced to 22 ℃ below the saturation temperature T2 corresponding to the steam generator pressure P2, wherein Tav is the average coolant temperature, P2 is the secondary side pressure of the steam generator at the broken pipe, T2 is the saturation temperature corresponding to the secondary side pressure of the steam generator, 22 ℃ is the supercooling degree standard value in the embodiment, the average coolant temperature meets the supercooling degree requirement, TP1 is 1, otherwise TP1 is 0;
according to logic and calculation results of an enabling signal Q, a main steam isolation valve state D1, an auxiliary water supply pump air supply valve state D2, a sewage pipeline valve state D3 and TP1, a change-over switch T is triggered together, wherein the T is a change-over switch and is used for being connected with a pressure set point measuring point of a pressure control system voltage stabilizer in a digital instrument control system of the nuclear power station, and when no heat transfer pipe fault occurs, namely, normal operation is carried out, a pressure reference value of the pressure control system keeps a pressure set value unchanged; when the breakage fault of the heat transfer pipe of the steam generator occurs, the switch T is switched to enable the pressure reference value of the pressure control system to be the secondary side pressure of the steam generator, the pressure reference value of the pressure control system is compared with the pressure P1 of the pressure stabilizer detected at the primary side in real time, namely the pressure of the pressure stabilizer changes along with the pressure of the steam generator, the pressure deviation between the pressure reference value and the pressure stabilizer is obtained, the opening state of the spraying valve of the pressure stabilizer is controlled through the pressure deviation, the spraying valve of the pressure stabilizer is sprayed, and automatic depressurization is realized;
the specific main program logic is as follows:
(e1) When u= "1" & s= "1", q= "1";
(e2) When |tp| < δ, tp1= "1";
when |tp| > δ, tp1= "0";
(e3)D1＝“1”，D2＝“1”，D3＝“1”；
(e4) When step (e 1), step (e 2) and step (e 3) are simultaneously satisfied, t= "1";
when t= "1", pref=p2;
when t= "0", pref=15.5 MPa;
(e5)TP＝Tav-(T2-22℃)，T2＝F(P2)，F(P2)＝F(x)；
F(x)＝6.535x 5 -0.004529x 4 +0.1325x 3 -2.194x 2 +26.33x 1 +173.4
wherein U is a heat transfer tube rupture signal; s is a shutdown signal; q is an enable signal; TP is the supercooling degree deviation; TP1 is a digital signal for judging whether the digital signal is within the margin of the supercooling degree; t is a change-over switch and is used for connecting a pressure reference value measuring point of a pressure control system voltage stabilizer in a digital instrument control system of the nuclear power station; pref is a reference value of the pressure control system, and 15.5MPa is a pressure set value of the pressure control system in the embodiment; d1 is a close main vapor isolation valve command; d2 is an instruction for closing an air supply valve of the steam-driven auxiliary water supply pump; d3 is a close blowdown valve command; tav is the coolant average temperature; t2 is the corresponding saturation temperature under the secondary side pressure of the steam generator; in this embodiment, the supercooling degree requirement of the coolant average temperature is: the average temperature of the coolant is reduced to 22 ℃ below the saturation temperature corresponding to the secondary side pressure of the steam generator, namely 22 ℃ is a supercooling standard value required for cooling during emergency operation of the nuclear power station, and as shown in figure 3, 22 ℃ is sent out by a given value signal generator;
f (P2) is a relation function between the secondary side pressure of the steam generator and the corresponding saturation temperature, and because the normal working pressure of the voltage stabilizer in the embodiment is 15.4Mpa and the normal working pressure of the steam generator is 6.51Mpa, F (x) is obtained by fitting the comparison data (shown in table 1) between the pressure of the steam generator and the saturation temperature in the pressure range of 6.5MPa to 16 MPa;
TABLE 1
As shown in fig. 3, when |tp| is within the δ range, TP 1=1 is output by the dead zone design, which is a temperature range required to detect whether the coolant temperature reaches the depressurization. When the |TP| is out of the delta range, the output is formed by Tp1=0, when the output is formed by Tp1=1, the output signal is combined with an enabling signal Q phase, the output signal 1 triggers a change-over switch T, pref=P2, the pressure reference value of the pressure control system is the secondary side pressure of the steam generator, and then the secondary side pressure of the steam generator is compared with the pressure of the voltage stabilizer;
when the output is TP 1=0, the output is taken as a Q phase of an enabling signal and a signal 0, at this time, the switch is switched to enable pref=15.5 Mpa, that is, the original pressure set value is maintained unchanged, and then 15.5Mpa is compared with the detected pressure value of the voltage stabilizer, in the embodiment, delta takes 0.2Mpa as a supercooling degree fluctuation margin.
(3) The full-automatic depressurization system in emergency treatment after the steam generator heat transfer pipe in the NETCONTROL system breaks down is communicated with the nuclear power station instrument control system through OPC.
And (3) performing point-to-point communication connection on the nuclear power station instrument control system and the NETCONTROL system through OPC. After communication connection, a program which is converted and edited according to a logic diagram in the NETCONTROL system is used for realizing full-automatic depressurization in emergency treatment after the steam generator heat transfer tube breakage accident:
when the steps (e 1) (e 2) (e 3) are simultaneously satisfied, the change-over switch t= "1", the real-time steam generator pressure P2 is used as the pressure reference value Pref of the pressure control system of the nuclear power plant, the pressure regulator pressure P1 and the pressure reference value Pref of the pressure control system are compared by a comparator to obtain pressure deviation (P1-Pref), and then calculated by a PID controller, and the output signal is the compensation pressure difference (P1-Pref) Tonifying device . When compensating for pressure differences (P1-Pref) Tonifying device When the pressure is in the range of 0.17-0.52 MPa, the opening of the spray valves 01VP and 02VP linearly changes along with the compensation pressure difference; when the compensation pressure difference is more than or equal to 0.52MPa, the valve is fully opened, and the spray valve is fully opened for the maximum spray flowRapid depressurization is performed.
When the steps (e 1) (e 2) (e 3) are not satisfied at the same time, the change-over switch t= "0", the pressure reference value (Pref) of the pressure control system of the nuclear power plant is not updated at the moment, the pressure reference value (Pref) is maintained to be the original pressure set value (the pressure set value is 15.5MPa in the embodiment) in the pressure control system, the pressure regulator pressure P1 and the pressure reference value Pref of the pressure control system are compared by a comparator to obtain pressure deviation (P1-Pref) (at the moment, pref is 15.5 MPa), and then the pressure deviation is calculated by a PID controller, and the output signal is the compensation pressure difference (P1-Pref) Tonifying device . When compensating for pressure differences (P1-Pref) Tonifying device When the pressure is in the range of 0.17-0.52 MPa, the opening of the spray valves 01VP and 02VP linearly changes along with the compensation pressure difference; when the compensation pressure difference is more than or equal to 0.52MPa, the valve is fully opened, and the maximum spraying flow is obtained.
In addition, when the power is 0-100% FP (i.e. the power under normal operation), if the corresponding compensation pressure difference (P1-Pref) Tonifying device In the range of 0.1 to-0.1 MPa, the proportional electric heaters 03RS and 04RS are controlled to heat the proportional electric heaters, and the power of the proportional electric heaters is controlled by the function generator 401 GD. When compensating for pressure differences (P1-Pref) Tonifying device When the pressure is reduced to-0.17 MPa, the threshold relay 430XU1 is used for switching on the on-off type electric heaters 01RS,02RS,05RS and 06RS and simultaneously starting to stop polarization operation; when compensating for pressure differences (P1-Pref) Tonifying device When the pressure rises to-0.1 MPa, the on-off electric heater is turned off.
The pressure alarm and protection measuring signal of the voltage stabilizer is from 015MP differential pressure gauge, and according to the comparison result of the measured value and alarm and protection setting value, a logic circuit and a relay generate proper logic signals to drive corresponding alarm or protection action. When the pressure of the voltage stabilizer is reduced to 15.2MPa, (L1) generating an alarm signal of low pressure of the voltage stabilizer; when the pressure of the voltage stabilizer is reduced to 14.9MPa (L2), the spray valves 01VP and 02VP are closed and the spray polarization operation is stopped; when the regulator pressure rises to 16.1MPa (H1), a signal is generated to close the regulator release tube scavenging valve RCP111 VY.
In summary, after the method provided by the invention is applied, the embodiment has the following advantages:
(1) After the heat transfer tube of the steam generator is broken, automatic depressurization is realized, the temperature is reduced firstly, the average temperature of the coolant system reaches 22 ℃ below the saturation temperature of the isolated steam generator, then depressurization is carried out, the automatic operation is timely, and excessive radioactive release can be effectively avoided.
(2) The automatic depressurization is triggered together based on fault diagnosis, shutdown and cooling, is rapid and timely and reliable, can reach the condition of stopping nuclear leakage in time, and has high reference value.
(3) The pressure control system of the actual nuclear power station is improved, and the change-over switch is additionally arranged, so that the proper pressure reference value connection point can be automatically switched and selected according to faults and normal operation conditions, the practicability is high, and the actual operability is strong.
Claims (3)
1. A nuclear power station primary loop depressurization control method under a heat transfer pipe rupture accident is characterized by comprising the following steps:
s1, acquiring fault information of a nuclear power station to calculate an enabling signal;
s2, acquiring the secondary side pressure of the steam generator and the average temperature of the coolant, and judging whether the average temperature of the coolant meets the supercooling degree requirement or not to obtain a cooling judgment result;
s3, acquiring isolation information of the steam generator, and calculating to obtain an update signal by combining an enabling signal and a cooling judgment result;
s4, obtaining a pressure reference value of the pressure control system based on the updating signal according to the secondary side pressure of the steam generator and the pressure set value of the pressure control system;
s5, obtaining the pressure of the voltage stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve so as to reduce the pressure of a loop;
the fault information of the nuclear power plant in the step S1 comprises a heat transfer pipe rupture signal and a shutdown signal, wherein the enabling signal is obtained by carrying out logic AND calculation on the heat transfer pipe rupture signal and the shutdown signal, when the shutdown signal and the heat transfer pipe rupture signal are both 1, the enabling signal is 1, otherwise, the enabling signal is 0;
the step S2 specifically includes the following steps:
s21, acquiring a secondary side pressure of the steam generator, and calculating to obtain the saturation temperature of the steam generator according to a relation function of the pressure and the saturation temperature of the steam generator;
s22, acquiring the average temperature of the coolant, and calculating the cooling degree deviation by combining the saturation temperature of the steam generator;
s23, judging whether the cooling degree deviation meets the requirement:
-δ＜TP＜+δ
wherein TP represents supercooling degree deviation, and delta represents supercooling degree fluctuation margin;
s24, if the supercooling degree deviation TP meets the requirement in the step S23, the average temperature of the coolant reaches the supercooling degree requirement to obtain a cooling judgment result of '1', otherwise, the average temperature of the coolant does not reach the supercooling degree requirement to obtain a cooling judgment result of '0';
in the step S3, the update signal is obtained by performing logic and calculation on the steam generator isolation information, the enable signal and the cooling judgment result, when the steam generator isolation information, the enable signal and the cooling judgment result are all "1", the update signal is "1", otherwise, the update signal is "0";
the steam generator isolation information in the step S3 includes a main steam isolation valve closing signal, a steam-driven auxiliary water supply pump steam supply valve closing signal, and a drain valve closing signal, wherein the signal value of the steam generator isolation information is obtained by performing logic and calculation on the main steam isolation valve closing signal, the steam-driven auxiliary water supply pump steam supply valve closing signal, and the drain valve closing signal, when the main steam isolation valve closing signal, the steam-driven auxiliary water supply pump steam supply valve closing signal, and the drain valve closing signal are all "1", the steam generator isolation information is "1", otherwise the steam generator isolation information is "0";
the step S4 specifically includes the following steps:
s41, taking the update signal as a switching condition, executing a step S42 when the update signal is 1, otherwise executing a step S43;
s42, setting the secondary side pressure value of the steam generator as a pressure reference value of a pressure control system;
s43, setting the pressure set value of the pressure control system as the pressure reference value of the pressure control system, namely maintaining the pressure set value of the current pressure control system unchanged.
2. The method for controlling the primary loop depressurization of the nuclear power plant in the event of a broken heat transfer pipe according to claim 1, wherein the calculation formula of the cooling deviation in step S22 is as follows:
TP＝Tav-(T2-Ts)
T2＝F(P2)
wherein Tav represents the average temperature of the coolant, T2 represents the saturation temperature of the failed steam generator, ts represents the standard value of the supercooling degree, P2 represents the secondary side pressure of the steam generator, F (x) is a function of the relationship between the steam generator pressure and the saturation temperature, x is the steam generator pressure, and F (P2) is a function of the relationship between the secondary side pressure of the steam generator and the saturation temperature.
3. The method for controlling the depressurization of the primary circuit of the nuclear power plant in the event of a broken heat transfer pipe according to claim 1, wherein said step S5 comprises the steps of:
s51, obtaining the pressure of the voltage stabilizer, and then performing difference calculation on the pressure of the voltage stabilizer and a pressure reference value of a pressure control system to obtain pressure deviation;
s52, controlling the opening value of the spraying valve of the voltage stabilizer through PID according to the pressure deviation, enabling the spraying valve to spray, and reducing the pressure of the primary loop.
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| CN111781075B (en) * | 2020-07-10 | 2021-05-07 | 西安交通大学 | Device and method for mechanical impact and vapor bubble migration experiment of lead-based stack evaporator heat transfer pipe fracture |
| CN113421663B (en) * | 2021-06-18 | 2022-04-15 | 中国核动力研究设计院 | Natural circulation cooling method suitable for pressurized water reactor nuclear power plant |
| CN115331858A (en) * | 2022-08-16 | 2022-11-11 | 中国核动力研究设计院 | Method for processing SGTR (steam generator and turbine control unit) accident of pressurized water reactor nuclear power plant and control system |
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