CN110689973A - Nuclear power station primary circuit pressure reduction control method under heat transfer pipe fracture accident - Google Patents
Nuclear power station primary circuit pressure reduction control method under heat transfer pipe fracture accident Download PDFInfo
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- 238000012546 transfer Methods 0.000 title claims abstract description 42
- 230000009467 reduction Effects 0.000 title claims abstract description 28
- 238000000034 method Methods 0.000 title claims abstract description 27
- 238000001816 cooling Methods 0.000 claims abstract description 26
- 238000002955 isolation Methods 0.000 claims abstract description 26
- 239000002826 coolant Substances 0.000 claims abstract description 25
- 238000004781 supercooling Methods 0.000 claims abstract description 24
- 239000003381 stabilizer Substances 0.000 claims abstract description 23
- 239000007921 spray Substances 0.000 claims abstract description 19
- 238000004364 calculation method Methods 0.000 claims abstract description 14
- 238000012544 monitoring process Methods 0.000 abstract description 7
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- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 3
- 230000010287 polarization Effects 0.000 description 2
- 238000005507 spraying Methods 0.000 description 2
- 230000009286 beneficial effect Effects 0.000 description 1
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- 238000013461 design Methods 0.000 description 1
- 238000001514 detection method Methods 0.000 description 1
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- 230000002000 scavenging effect Effects 0.000 description 1
- 239000010865 sewage Substances 0.000 description 1
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C9/00—Emergency protection arrangements structurally associated with the reactor, e.g. safety valves provided with pressure equalisation devices
- G21C9/004—Pressure suppression
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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
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
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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
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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Abstract
The invention relates to a nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident, which comprises the following steps: s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation; s2, judging whether the average temperature of the coolant meets the supercooling degree requirement or not to obtain a cooling judgment result; s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result; s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value; and S5, acquiring 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 as to reduce the pressure of the loop. 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 judgment time are solved, and the accuracy and the speed of emergency treatment are improved.
Description
Technical Field
The invention relates to the technical field of nuclear power station steam generator heat transfer pipe rupture accident treatment, in particular to a nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident.
Background
The rupture of the heat transfer pipe of the steam generator of the nuclear power plant refers to the rupture of one or more heat transfer pipes in the steam generator, so that the pressure boundary of a primary loop loses integrity, the primary loop is communicated with a secondary loop, and a nuclear leakage accident is caused. For this reason, the existing emergency treatment method is usually that after a heat transfer pipe rupture accident occurs, an operator enters a standard operation rule according to the requirements of a corresponding alarm or technical specification program to perform accident control, and the main control strategy is as follows:
1. identifying and isolating a faulty steam generator;
2. under the condition of ensuring the supercooling degree, controlling a loop to reduce the pressure as soon as possible, and reducing and eliminating leakage;
3. when the pressure of the first loop and the pressure of 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 above process requires that the operator continuously judges the master control state and then controls the accident by adopting corresponding means, but in the actual manual operation process, the judgment operation time is long due to the complex steps, so that the manual intervention measures are not timely and accurate enough, and especially when the manual pressure reduction of a loop is carried out, the operation difficulty is further increased due to the excessive conditions needing 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 in the prior art and provide a primary circuit pressure reduction control method for a nuclear power station in the event of a heat transfer pipe rupture accident.
The purpose of the invention can be realized by the following technical scheme: a nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident comprises the following steps:
s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation;
s2, acquiring the secondary 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 to obtain a cooling judgment result;
s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result;
s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value;
and S5, acquiring 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 as to reduce the pressure of the loop.
Further, the nuclear power plant fault information in step S1 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 both the shutdown signal and the heat transfer pipe rupture signal are "1", the enable signal is "1", and if not, the enable signal is "0".
Further, the step S2 specifically includes the following steps:
s21, acquiring secondary pressure of the steam generator, and calculating to obtain saturation temperature of the steam generator according to a relation function of the pressure and the saturation temperature of the steam generator;
s22, obtaining the average temperature of the coolant, and calculating the deviation of the cooling degree 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 meets the supercooling degree requirement, the obtained cooling judgment result is '1', otherwise, the average temperature of the cooling degree does not meet the supercooling degree requirement, and the obtained cooling judgment result is '0'.
Further, the calculation formula of the cooling degree deviation TP in the step S22 is:
TP＝Tav-(T2-Ts)
T2＝F(P2)
wherein Tav represents the average coolant temperature, T2 represents the saturation temperature of the failed steam generator, Ts represents the subcooling threshold value, P2 represents the secondary steam generator pressure, F (x) is the steam generator pressure as a function of the saturation temperature, x is the steam generator pressure, and F (P2) is the secondary steam generator pressure as a function of the saturation temperature.
Further, the update signal in step S3 is obtained by performing logic and calculation on the steam generator isolation information, the enable signal, and the cooling determination result, where the update signal is "1" when the steam generator isolation information, the enable signal, and the cooling determination result are all "1", and the update signal is "0" otherwise.
Further, in step S3, the steam generator isolation information includes a main steam isolation valve closing signal, a steam-driven auxiliary water-feeding pump steam-feeding valve closing signal, and a blowoff valve closing signal, where a signal value of the steam generator isolation information is obtained by performing logical and calculation on the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam-feeding valve closing signal, and the blowoff valve closing signal, and when the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam-feeding valve closing signal, and the blowoff 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 the switching condition, executing step S42 when the update signal is '1', otherwise executing step S43;
s42, setting the pressure value of the secondary side of the steam generator as a pressure control system pressure reference value;
and 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 pressure stabilizer, and then performing difference calculation on the pressure of the pressure stabilizer and a pressure reference value of a pressure control system to obtain pressure deviation;
and S52, controlling the opening value of the spray valve of the voltage stabilizer through the PID control according to the pressure deviation to spray the spray valve, and reducing the pressure of the primary loop.
Compared with the prior art, the invention has the following advantages:
the invention uses the logic and result of the heat transfer pipe rupture signal and the shutdown signal as the enabling signal, combines the evaporator isolation information and the cooling judgment result, and uses the logic and result as the updating signal for switching and setting the pressure reference value of the pressure control system, thereby being capable of automatically reducing the pressure after ensuring effective cooling, rapidly and timely reducing the pressure and reliably meeting the requirement of stopping nuclear leakage.
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 judgment time, is beneficial to the follow-up timely control of the opening value of the spray valve by the pressure control system, does not influence the normal operation of the existing pressure control system, and has strong practical operability.
Drawings
FIG. 1 is a flow chart of a method of the present invention;
FIG. 2 is a schematic diagram of a system for implementing a voltage reduction control method in an embodiment;
fig. 3 is a logic structure diagram of the voltage reduction control method in the embodiment.
The notation in the figure is:
Detailed Description
The invention is described in detail below with reference to the figures and specific embodiments.
As shown in fig. 1, a method for controlling the primary circuit depressurization of a nuclear power plant in the event of a heat transfer pipe rupture comprises the following steps:
s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation;
s2, acquiring the secondary 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 to obtain a cooling judgment result;
s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result;
s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value;
and S5, acquiring 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 as to reduce the pressure of the loop.
As shown in fig. 2, in this embodiment, monitoring data is read by a nuclear power plant instrumentation and control system, a program of a nuclear power plant primary circuit automatic voltage reduction control logic is compiled after a heat transfer pipe rupture accident is performed on a NETCONTROL system platform, mutual communication between the NETCONTROL system program and the nuclear power plant instrumentation and control system is realized by using OPC communication, specifically, an OPC is used as a communication tool, a variable database corresponding to an acquired monitoring point is established in the NETCONTROL system, data acquired from the nuclear power plant instrumentation and control system is set as an input variable group, data output to the nuclear power plant instrumentation and control system is set as an output variable group, and a variable calculated inside the NETCONTROL system is set as an intermediate variable group. On the basis of setting the finished variable group, editing a script program in a NETCONTROL system according to the logical relation of the automatic voltage reduction control method, and finally realizing automatic voltage reduction control in emergency response after a steam generator heat transfer pipe breakage accident.
The logic structure of the voltage reduction control method in the embodiment is shown in fig. 3, and the specific application process includes:
(1) monitoring a fault diagnosis system and a nuclear power station instrument control system in real time on line, and acquiring the state and monitoring data of monitoring points of the nuclear power station:
respectively acquiring a shutdown signal S, a pressure P1 of a voltage stabilizer, a secondary pressure P2 of a steam generator, an average temperature Tav of a coolant, a state D1 of a main steam isolation valve, a state D2 of an air supply valve of an auxiliary water feed pump and a state D3 of a drain pipeline valve 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) The automatic voltage reduction control method based on the emergency treatment after the steam generator heat transfer pipe rupture accident is realized in a NETCONTROL system. When the steam generator heat transfer tube is broken, the shutdown signal S sent by the protection system and the steam generator heat transfer tube breakage signal U sent by the fault diagnosis system are obtained, and the two signals are in phase with each other to obtain an enabling signal Q, when the steam generator heat transfer tube is broken, 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, the pressure reduction operation is automatically triggered, and when the enable signal Q is '1', the coolant system is triggered to automatically reduce the average coolant temperature Tav, so that the temperature Tav of the coolant system is reduced to 22 ℃ below the saturation temperature T2 corresponding to the pressure P2 of the steam generator, wherein the Tav is the average coolant temperature, P2 is the secondary pressure of the steam generator at the pipe breakage position, T2 is the saturation temperature standard value corresponding to the secondary 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', and otherwise TP1 is '0';
according to the logic and calculation results of an enabling signal Q, a main steam isolation valve state D1, an auxiliary water feeding pump air supply valve state D2, a sewage pipeline valve state D3 and a TP1, a change-over switch T is triggered together and used for being connected with a pressure set value test 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 maintains the pressure set value unchanged; when the steam generator heat transfer pipe breaks and fails, the pressure reference value of the pressure control system is the secondary pressure of the steam generator by the change-over switch T, the pressure reference value of the pressure control system is compared with the pressure P1 of the voltage stabilizer detected on the primary side in real time, namely, the pressure of the voltage stabilizer is changed along with the pressure of the steam generator to obtain the pressure deviation between the pressure reference value and the pressure stabilizer, and the opening state of a spray valve of the voltage stabilizer is controlled by the pressure deviation to spray the spray valve of the voltage stabilizer, so that automatic pressure reduction is realized;
the specific main program logic is step e:
(e1) when U ═ 1 "& S ═ 1", Q ═ 1 ";
(e2) when | TP | < δ, TP1 ═ 1;
when | TP | > δ, TP1 is "0";
(e3)D1＝“1”，D2＝“1”，D3＝“1”；
(e4) when step (e1), step (e2) and step (e3) are simultaneously satisfied, T ═ 1 ";
when T is "1", Pref is P2;
when T is equal to 0', Pref is equal to 15.5 MPa;
(e5)TP＝Tav-(T2-22℃)，T2＝F(P2)，F(P2)＝F(x)；
F(x)＝6.535x5-0.004529x4+0.1325x3-2.194x2+26.33x1+173.4
wherein, U is a heat transfer pipe rupture signal; s is a shutdown signal; q is an enable signal; TP is supercooling degree deviation; TP1 is a digital signal for judging whether the supercooling degree is within the allowance of the supercooling degree; t is a change-over switch used for connecting a pressure reference value measuring point of a pressure control system voltage stabilizer in a nuclear power station digital instrument control system; 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 steam isolation valve command; d2 is an instruction for closing an air supply valve of the steam-driven auxiliary water feed pump; d3 is a close blowdown valve command; tav is the average coolant temperature; t2 is the corresponding saturation temperature under the secondary pressure of the steam generator; in this embodiment, the supercooling degree requirement of the average temperature of the coolant is as follows: the average temperature of the coolant is reduced to 22 ℃ below the corresponding saturation temperature of the secondary pressure of the steam generator, namely 22 ℃ is a supercooling standard value required by temperature reduction in emergency operation of the nuclear power station, and the temperature of 22 ℃ is sent by a set value signal generator as shown in figure 3;
f (P2) is a relation function of the secondary pressure of the steam generator and the corresponding saturation temperature, and since the pressure of the regulator in the embodiment is 15.4MPa and the pressure of the steam generator in the embodiment is 6.51MPa, F (x) in the embodiment is obtained by fitting the comparison data (shown in Table 1) of 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 in the δ range, the output is TP1 ═ 1 by the dead band design, which is to detect whether the coolant temperature reaches the temperature range required for pressure reduction. When | TP | is out of the δ range, the output is TP1 ═ 0, and when TP1 ═ 1, the output is anded with an enable signal Q, the output signal 1 triggers a change-over switch T, Pref ═ P2, the pressure control system pressure reference value is the steam generator secondary pressure, and then the steam generator secondary pressure is compared with the regulator pressure;
when the output is TP1 is equal to 0, the signal is anded with the enable signal Q, the signal 0 is output, at this time, the switch is switched so that Pref is equal to 15.5Mpa, 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, wherein in the embodiment, the value delta is 0.2Mpa, which is the supercooling degree fluctuation margin.
(3) The full-automatic depressurization system and the nuclear power station instrument control system in emergency treatment after the steam generator heat transfer pipe in the NETCONTROL system is broken are communicated with each other through OPC.
And carrying out point-to-point communication connection on the instrument control system of the nuclear power station and the NETCONTROL system through OPC. After communication connection, the full-automatic voltage reduction based on the steam generator heat transfer pipe rupture accident in the emergency treatment is realized by using a program which is converted and edited according to a logic diagram in a NETCONTROL system:
when the steps (e1) (e2) (e3) are simultaneously satisfied, the change-over switch T is equal to '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 at the time through the change-over switch T, the pressure regulator pressure P1 and the pressure control system pressure reference value Pref are compared through a comparator to obtain a pressure deviation (P1-Pref), then the pressure deviation is calculated through a PID controller, and an output signal is a compensation pressure difference (P1-Pref)Supplement device. When compensating for the pressure difference (P1-Pref)Supplement deviceWhen the pressure is within the range of 0.17-0.52 MPa, the opening degrees of the spray valves 01VP and 02VP are linearly changed along with the compensation pressure difference; when the compensating pressure difference is more than or equal to 0.52MPa, the valve is fully opened and the maximum spraying is realizedAnd (4) rapidly reducing the pressure by fully opening the spray valve.
When the steps (e1) (e2) (e3) are not simultaneously satisfied, the transfer switch T is equal to "0", the pressure reference value (Pref) of the pressure control system of the nuclear power plant is not updated through the transfer switch T and is maintained as the original pressure set value in the pressure control system (in the embodiment, the pressure set value is 15.5MPa), the pressure regulator pressure P1 and the pressure reference value Pref of the pressure control system are compared through a comparator to obtain a pressure deviation (P1-Pref) (the Pref is 15.5MPa), and then the pressure deviation is calculated through a PID controller, and an output signal is a compensation pressure difference (P1-Pref)Supplement device. When compensating for the pressure difference (P1-Pref)Supplement deviceWhen the pressure is within the range of 0.17-0.52 MPa, the opening degrees of the spray valves 01VP and 02VP are linearly changed along with the compensation pressure difference; when the compensating pressure difference is more than or equal to 0.52MPa, the valve is fully opened, and the maximum spraying flow is achieved.
In addition, when the power is 0-100% FP (namely the power under the normal operation condition), if the corresponding compensation pressure difference (P1-Pref)Supplement deviceIn 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 a function generator 401 GD. When compensating for the pressure difference (P1-Pref)Supplement deviceWhen the pressure is reduced to-0.17 MPa, the on-off electric heaters 01RS, 02RS, 05RS and 06RS are switched on simultaneously by the threshold relay 430XU1, and the polarization operation is stopped; when compensating for the pressure difference (P1-Pref)Supplement deviceAnd when the pressure rises to-0.1 MPa, the on-off electric heater is closed.
The pressure alarm and protection measuring signal of the voltage stabilizer comes from a 015MP differential pressure meter, and a logic circuit and a relay generate appropriate logic signals to drive corresponding alarm or protection actions according to the comparison result of the measured value and the alarm and protection setting value. When the pressure of the voltage stabilizer is reduced to 15.2MPa (L1), an alarm signal of 'low pressure of the voltage stabilizer' is generated; 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 pressurizer pressure rises to 16.1MPa (H1), a signal is generated to close the pressurizer release line scavenging valve RCP111 VY.
In summary, the method of the present invention has the following advantages:
(1) the automatic pressure reduction is realized after the heat transfer pipe of the steam generator is broken, the temperature is firstly reduced, the average temperature of a coolant system reaches 22 ℃ below the saturation temperature of the isolated steam generator, then the pressure reduction is carried out, the automatic operation is timely, and the release of excessive radioactivity can be effectively avoided.
(2) The automatic depressurization is triggered based on fault diagnosis, shutdown and temperature reduction, is rapid, timely and reliable, can timely achieve the condition of stopping nuclear leakage, and has high reference value.
(3) The pressure control system of the nuclear power station is improved, the change-over switch is additionally arranged, the proper pressure reference value connecting point can be automatically switched and selected according to the fault and normal operation conditions, the practicability is high, and the actual operability is strong.
Claims (8)
1. A nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident is characterized by comprising the following steps:
s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation;
s2, acquiring the secondary 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 to obtain a cooling judgment result;
s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result;
s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value;
and S5, acquiring 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 as to reduce the pressure of the loop.
2. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture accident according to claim 1, wherein the nuclear power plant fault information in step S1 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 both the shutdown signal and the heat transfer pipe rupture signal are "1", the enable signal is "1", and if not, the enable signal is "0".
3. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture as recited in claim 1, wherein said step S2 specifically comprises the steps of:
s21, acquiring secondary pressure of the steam generator, and calculating to obtain saturation temperature of the steam generator according to a relation function of the pressure and the saturation temperature of the steam generator;
s22, obtaining the average temperature of the coolant, and calculating the deviation of the cooling degree 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 meets the supercooling degree requirement, the obtained cooling judgment result is '1', otherwise, the average temperature of the cooling degree does not meet the supercooling degree requirement, and the obtained cooling judgment result is '0'.
4. A primary circuit pressure reducing control method for a nuclear power plant in the event of a heat transfer pipe rupture as set forth in claim 3, wherein the calculation formula of the cooling deviation TP in step S22 is:
TP＝Tav-(T2-Ts)
T2＝F(P2)
wherein Tav represents the average coolant temperature, T2 represents the saturation temperature of the failed steam generator, Ts represents the subcooling threshold value, P2 represents the secondary steam generator pressure, F (x) is the steam generator pressure as a function of the saturation temperature, x is the steam generator pressure, and F (P2) is the secondary steam generator pressure as a function of the saturation temperature.
5. The method for controlling the primary circuit pressure drop of the nuclear power plant in the event of a heat transfer tube rupture accident as recited in claim 1, wherein the update signal in step S3 is obtained by logically and-calculating the steam generator isolation information, the enable signal and the cooling determination result, and the update signal is "1" when the steam generator isolation information, the enable signal and the cooling determination result are all "1", and otherwise the update signal is "0".
6. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture accident according to claim 1, wherein the steam generator isolation information in step S3 includes a main steam isolation valve closing signal, a steam-driven auxiliary water-feeding pump steam supply valve closing signal, and a blow-off valve closing signal, wherein the signal values of the steam generator isolation information are obtained by performing logical and calculation on the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam supply valve closing signal, and the blow-off valve closing signal, and when the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam supply valve closing signal, and the blow-off valve closing signal are all "1", the steam generator isolation information is "1", otherwise, the steam generator isolation information is "0".
7. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture as recited in claim 1, wherein said step S4 specifically comprises the steps of:
s41, taking the update signal as the switching condition, executing step S42 when the update signal is '1', otherwise executing step S43;
s42, setting the pressure value of the secondary side of the steam generator as a pressure control system pressure reference value;
and 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.
8. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture as recited in claim 1, wherein said step S5 specifically comprises the steps of:
s51, obtaining the pressure of the pressure stabilizer, and then performing difference calculation on the pressure of the pressure stabilizer and a pressure reference value of a pressure control system to obtain pressure deviation;
and S52, controlling the opening value of the spray valve of the voltage stabilizer through the PID control according to the pressure deviation to spray the spray valve, and reducing the pressure of the primary loop.
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| CN111540487A (en) * | 2020-04-30 | 2020-08-14 | 中国核动力研究设计院 | Cooling treatment method for reactor after steam generator heat transfer pipe failure accident |
| CN111781075A (en) * | 2020-07-10 | 2020-10-16 | 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 |
| CN113421663A (en) * | 2021-06-18 | 2021-09-21 | 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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| CN113421663A (en) * | 2021-06-18 | 2021-09-21 | 中国核动力研究设计院 | Natural circulation cooling method suitable for pressurized water reactor nuclear power plant |
| 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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