CN111681794A - Full-range SGTR accident handling method and system for pressurized water reactor nuclear power plant - Google Patents
Full-range SGTR accident handling method and system for pressurized water reactor nuclear power plant Download PDFInfo
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Abstract
The invention discloses a full-range SGTR accident handling method and system for a pressurized water reactor nuclear power plant, which comprises the following steps: s1: judging that the accident of steam generator heat transfer tube rupture occurs according to the narrow range water level of the steam generator and the high signal of the radioactivity of the two loops; s2: judging the type of the steam generator heat transfer pipe rupture accident after the steam generator heat transfer pipe rupture accident occurs; s3: and adopting a corresponding accident handling method according to the type of the steam generator heat transfer pipe breakage accident. According to the characteristics of the steam generator heat transfer pipe rupture accidents, the steam generator heat transfer pipe rupture accidents are quickly and effectively treated by arranging the corresponding accident treatment method, the radioactive release to the environment can be reduced, and the response range of the steam generator heat transfer pipe rupture accident treatment strategy is greatly expanded.
Description
Technical Field
The invention relates to a method for handling accidents of cracking of a steam generator heat transfer pipe of a pressurized water reactor nuclear power plant (station), in particular to a method and a system for handling all-range SGTR (steam generator controller) accidents of the pressurized water reactor nuclear power plant.
Background
After a steam generator heat transfer pipe rupture accident occurs in a pressurized water reactor nuclear power plant (station), the radioactivity of a primary loop is released into a secondary loop system through a rupture, and then is released into the environment through a secondary loop atmospheric release valve/safety valve. Therefore, operators need to take corresponding actions to reduce the radioactive emission of the broken steam generator according to the countermeasures of the steam generator heat transfer pipe breakage accident. The strategy for dealing with the steam generator heat transfer pipe rupture accidents in China is mainly used for dealing with the steam generator heat transfer pipe rupture accidents (steam generator heat transfer pipe rupture accidents, SGTR) of general types, and the strategy cannot deal with the steam generator heat transfer pipe rupture type superposition accidents (such as primary loop water loss due to SGTR superposition, primary loop pressure runaway due to SGTR superposition, secondary loop faults due to SGTR superposition and the like).
Disclosure of Invention
The invention aims to solve the technical problems that the prior art is mainly used for processing general steam generator heat transfer pipe rupture accidents and cannot process steam generator heat transfer pipe rupture overlapping accidents, and aims to provide a method and a system for processing the full-range SGTR accidents of a pressurized water reactor nuclear power plant, so as to solve the problem of safely, effectively and quickly processing the general steam generator heat transfer pipe rupture accidents and the overlapping accidents thereof, namely, safely, effectively and quickly processing the full-range SGTR accidents.
The invention is realized by the following technical scheme:
a full-range SGTR accident handling method for a pressurized water reactor nuclear power plant comprises the following steps: s1: judging whether a steam generator heat transfer pipe rupture accident occurs or not according to the narrow-range water level of the steam generator and the two-loop radioactive signal; s2: judging the type of the steam generator heat transfer tube rupture accident after the steam generator heat transfer tube rupture accident occurs, wherein the type of the steam generator heat transfer tube rupture accident comprises the following steps: a first type of accident, a second type of accident, a third type of accident, and a fourth type of accident; the first type of accident is a general type steam generator heat transfer pipe rupture accident, the second type of accident is a steam generator heat transfer pipe rupture accident superimposed primary loop water loss accident, the third type of accident is a steam generator heat transfer pipe rupture accident superimposed primary loop pressure runaway, and the fourth type of accident is a steam generator heat transfer pipe rupture accident superimposed secondary side fault; s3: and according to the type of the steam generator heat transfer pipe breakage accident, using a corresponding accident handling step.
The method for handling the steam generator heat transfer pipe rupture accident in the prior art is mainly used for handling the steam generator heat transfer pipe rupture accident of a general type and cannot handle steam generator heat transfer pipe rupture superposition accidents (such as primary circuit water loss due to SGTR superposition, primary circuit pressure runaway due to SGTR superposition, secondary circuit faults due to SGTR superposition and the like). The method firstly judges the accident of the pressurized water reactor nuclear power plant (station), judges the type of the steam generator heat transfer pipe rupture accident when the accident is really the steam generator heat transfer pipe rupture accident, and takes corresponding measures according to the characteristics of the steam generator heat transfer pipe rupture accidents of different types. After the steam generator heat transfer pipe rupture accident happens, an operator judges the specific steam generator heat transfer pipe rupture accident type according to the actual response of a power plant, and reduces the radioactive release of the damaged steam generator through a reasonable accident handling method. The method expands the coping range of the handling strategy of the steam generator heat transfer pipe rupture accidents and optimizes the handling strategy of the steam generator heat transfer pipe rupture accidents.
Further, the step S2 includes: judging whether the steam generator heat transfer pipe breakage accident is a first accident or not according to the state of the secondary side of the intact steam generator; judging whether the steam generator heat transfer pipe breakage accident is a second accident or not according to the state of a loop system; judging whether the steam generator heat transfer pipe breakage accident is a third accident or not according to the condition whether main spray, auxiliary spray and a safety valve of the pressure stabilizer are unavailable or not; and judging whether the steam generator heat transfer pipe breakage accident is a fourth accident or not according to the state of the two-circuit system.
Because the types of the steam generator heat transfer pipe rupture accidents of the pressurized water reactor nuclear power plant (station) have respective characteristics, the types of the steam generator heat transfer pipe rupture accidents need to be judged according to different states of various devices so as to distinguish the types of the steam generator heat transfer pipe rupture accidents, and a reasonable and safe processing method is provided.
Further, the second type of accident comprises a steam generator heat transfer tube rupture accident superposition stabilizer safety valve clamping-opening accident, a single steam generator multi-heat transfer tube rupture accident and a steam generator heat transfer tube rupture accident superposition primary circuit breach accident.
Further, judging whether the safety valve of the superposed voltage stabilizer is blocked according to the state of the safety valve of the voltage stabilizer; and judging whether a primary loop break accident is superposed or a plurality of heat transfer tubes of a single steam generator break accident occurs according to whether the primary loop pressure is too low, the supercooling degree of the reactor core outlet is continuously reduced, the safety injection pump cannot be stopped, the safety injection pump needs to be restarted after being stopped or the safety injection tank cannot be isolated.
Further, the secondary side fault includes: the accident of two-loop pipeline breakage, the accident of steam generator safety valve or release valve blockage, the accident of damaged steam generator unable isolation and the accident of secondary side heat trap loss.
Further, judging whether the secondary side fault is a secondary circuit pipeline break accident or not according to the pressure of the damaged steam generator; judging whether the secondary side fault is a safety valve or release valve stuck-open accident or not according to the state of a safety valve or an atmospheric release valve of the damaged steam generator; judging whether the secondary side fault is a failure of isolating the damaged steam generator or not according to the state of an isolating valve of the damaged steam generator; and judging whether the secondary side fault is a secondary side heat trap loss accident or not according to the water supply flow of the damaged steam generator.
Further, the step S3 includes:
when a first type accident occurs, the accident handling steps are as follows: identifying and isolating a damaged steam generator, cooling a primary loop system to establish supercooling degree, reducing pressure of the primary loop to recover water content of the primary loop system, stopping safety injection to stop leakage from a primary side to a secondary side, and cooling the primary loop system to a cold shutdown working condition, and finally maintaining the nuclear power plant in the cold shutdown working condition;
when a second type of accident occurs, the accident handling steps are as follows: the primary loop system cools and reduces the pressure as soon as possible to reduce the loss of the coolant, reduces the flow of safety injection in time and stops the safety injection in time to reduce the possibility of overflow of a steam generator, realizes cold shutdown and primary loop system pressure reduction to stop leakage of radioactive substances according to the actual state of the power plant at the allowed fastest speed or the fastest speed which can be achieved by the power plant, puts into a normal waste heat discharge system, and finally brings the reactor device into a cold shutdown state;
when a third type of accident occurs, the accident handling steps are as follows: recovering pressure control of the voltage stabilizer, stopping safety injection to stop leakage of the primary side and the secondary side, preparing and starting to cool a loop system, putting into a normal waste heat discharge system, and finally bringing the reactor into a cold shutdown state;
when a fourth type of accident occurs, the accident handling steps are as follows: the primary loop system cools and reduces the pressure as soon as possible to reduce the loss of the coolant, reduces the flow of safety injection in time and stops the safety injection in time to reduce the possibility of overflow of the steam generator, realizes cold shutdown and primary loop system pressure reduction to stop leakage of radioactive substances at the allowed fastest speed or the fastest speed which can be achieved by the power plant according to the actual state of the power plant, puts into a normal waste heat discharge system, and finally brings the reactor device into a cold shutdown state.
Because the types of the steam generator heat transfer pipe rupture accidents of the pressurized water reactor nuclear power plant (station) have respective characteristics, the steam generator heat transfer pipe rupture accidents need to be dealt with rapidly and effectively according to the types of the steam generator heat transfer pipe rupture accidents, the radioactive release to the environment is reduced, and the safe operation of the pressurized water reactor nuclear power plant (station) is guaranteed.
Further, the first type of accident includes: a single steam generator tube burst event, two steam generator tube burst events, and multiple steam generator tube burst events.
Further, the two-loop radioactivity signal comprises: monitoring the leakage quantity N-16 of the steam generator, the total gamma activity of the steam, the gamma activity of the air extracted by the condenser and the gamma activity of the sewage discharged by the steam generator.
In another implementation manner of the present invention, a system for handling a full-scale SGTR accident in a pressurized water reactor nuclear power plant includes:
a monitoring module: the system is used for monitoring the states of all equipment in the pressurized water reactor nuclear power plant;
a judging module: the system is used for judging whether a steam generator heat transfer pipe rupture accident occurs or not according to the states of all equipment in the pressurized water reactor nuclear power plant, judging whether the steam generator heat transfer pipe rupture accident has a superposition accident or not and judging the type of the steam generator heat transfer pipe rupture accident;
a processing module: the accident handling method comprises the steps of establishing a corresponding accident handling process according to the type of the steam generator heat transfer pipe breakage accident;
a display module: and the accident type and the processing flow thereof are displayed.
Compared with the prior art, the invention has the following advantages and beneficial effects:
1. according to the method for handling the steam generator heat transfer pipe rupture accidents in the full range, disclosed by the invention, the steam generator heat transfer pipe rupture accidents are quickly and effectively handled by arranging the corresponding accident handling method according to the characteristics of the steam generator heat transfer pipe rupture accidents, the radioactive release to the environment can be reduced, and the coping range of the steam generator heat transfer pipe rupture accident handling strategy is greatly expanded.
2. The invention relates to an accident handling method after a full-range steam generator heat transfer pipe rupture accident (including a steam generator heat transfer pipe rupture accident and a primary loop or secondary loop accident superimposed on the steam generator heat transfer pipe rupture accident) of a pressurized water reactor nuclear power plant (station). After the steam generator heat transfer pipe rupture accident happens, an operator judges the specific steam generator heat transfer pipe rupture accident type according to the actual response of the power plant, and reduces the radioactive release of the damaged steam generator by setting a reasonable accident handling method.
3. The method expands the coping range of the handling strategy of the steam generator heat transfer pipe rupture accidents and optimizes the handling strategy of the steam generator heat transfer pipe rupture accidents.
4. Through the steam generator heat transfer pipe rupture accident judgment and operation key points, an operator can adopt the most appropriate accident handling strategy to handle the accident according to the full-range steam generator heat transfer pipe rupture accident handling method, and the radioactivity release after the accident is reduced.
5. The method is based on the accident characteristics under different working conditions of the steam generator heat transfer pipe rupture accident, the type of the steam generator heat transfer pipe rupture accident is rapidly judged, the optimal steam generator heat transfer pipe rupture accident coping scheme is selected, the method for processing the steam generator rupture accident is optimized, the international leading technical level is achieved, and the method has very important significance for improving the design technology of the current three-generation nuclear power plant (station) in China. The invention fills up the related blank field of the safety strategy design of the three generations of nuclear power plants (stations) in China, and has the potential of entering the military and international markets at the same time.
Drawings
The accompanying drawings, which are included to provide a further understanding of the embodiments of the invention and are incorporated in and constitute a part of this application, illustrate embodiment(s) of the invention and together with the description serve to explain the principles of the invention. In the drawings:
FIG. 1 is a flowchart of a processing method of example 2.
Detailed Description
In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit the present invention.
Example 1
In this embodiment 1, from the beginning of the event, the steam generator tube rupture accident of the pressurized water reactor nuclear power plant (station) mainly includes the following operating conditions:
1. steam generator heat transfer tube rupture accidents of the general type: the method comprises the following steps of a single steam generator heat transfer pipe rupture accident, two steam generator heat transfer pipe rupture accidents, a plurality of steam generator heat transfer pipe rupture accidents and the like.
2. Superposition of steam generator heat transfer tube rupture accident and primary loop water loss accident: the method comprises the accident of blocking and unlocking of the safety valve of the SGTR (steam generator controller) superposition voltage stabilizer, the accident of rupture of a plurality of heat transfer pipes of a single steam generator and the accident of breakage of an SGTR superposition primary circuit.
3. Superposition of steam generator heat transfer tube rupture accident and primary loop pressure out of control: including the accident of pressure runaway of an SGTR (steam generator and gas turbine) superposed primary circuit and the like.
4. Superposition of secondary side faults due to steam generator heat transfer pipe rupture accidents: the accident that the steam generator cannot be isolated by the SGTR overlapping damage, the accident that the atmosphere release valve or the atmosphere release valve of the SGTR overlapping steam generator is blocked, the accident that the SGTR overlapping two-loop break, the accident that the SGTR overlapping intact steam generator cannot be used and the like are included.
The most obvious accident characteristic of the steam generator heat transfer tube breakage is that the leakage of primary loop coolant to the secondary loop causes the radioactivity of the secondary loop and the water content of the secondary side of the steam generator to increase, and an operator can easily judge that the steam generator heat transfer tube breakage accident occurs according to the narrow range water level of the steam generator and the high signal of the radioactivity of the secondary loop. An operator can judge the specific type of the steam generator heat transfer pipe rupture accident according to the typical characteristics of each superposed accident, and make a full-range steam generator heat transfer pipe rupture accident coping strategy, such as:
1. steam generator heat transfer tube rupture accidents of the general type: and judging whether the accident of the rupture of the heat transfer tubes of the plurality of steam generators occurs or not according to the secondary side state (uncontrollable rising of SG water level) of the 'intact' steam generator.
For a steam generator heat transfer tube rupture accident of a general type, the operator's operational points include: the method comprises the steps of identifying and isolating a damaged steam generator, cooling a primary loop system to establish supercooling degree, reducing pressure of the primary loop to recover water content of the primary loop system, stopping safety injection to stop leakage of a primary side and a secondary side, cooling the primary loop system to a cold shutdown working condition and the like, and finally maintaining the nuclear power plant in the cold shutdown working condition.
2. Superposition of steam generator heat transfer tube rupture accident and primary loop water loss accident: according to a loop system state during the SGTR accident, can judge that the SGTR superposes a loop loss of coolant accident, include: judging whether the safety valve of the superposed voltage stabilizer is blocked according to the state of the safety valve of the voltage stabilizer; and judging whether a primary circuit break accident or a single steam generator multi-heat-transfer-pipe break accident is superposed according to the conditions of whether the primary circuit pressure is too low, or the supercooling degree of the reactor core outlet is continuously reduced, or the safety injection pump cannot be stopped, or the safety injection pump needs to be restarted after being stopped, or the safety injection tank cannot be isolated, and the like.
To steam generator heat-transfer pipe rupture accident stack a return circuit water loss accident, operator's operating key points include: the primary loop system is cooled and depressurized as soon as possible to reduce loss of coolant, safety injection flow is reduced in time, safety injection is stopped in time to reduce possibility of overflow of a steam generator, cold shutdown is realized at the fastest allowable speed or the fastest speed which can be achieved by the power plant according to actual states of the power plant, the primary loop system is depressurized to stop leakage of radioactive substances, a normal surplus discharge system is put into use, and finally the reactor device is brought into a cold shutdown state.
3. Superposition of steam generator heat transfer tube rupture accident and primary loop pressure out of control: and judging whether the primary loop pressure is out of control or not by superposing the steam generator heat transfer pipe rupture accidents according to the fact that the main spray, the auxiliary spray and the safety valve of the pressure stabilizer are unavailable.
To steam generator heat-transfer pipe rupture accident stack return circuit pressure out of control, operator's operating essential includes: recovering the pressure control of the voltage stabilizer, stopping safety injection to stop leakage from the primary side to the secondary side, preparing and starting to cool a primary loop system, putting into normal residual discharge, and finally bringing the reactor into a cold shutdown state.
4. Superposition of secondary side faults due to steam generator heat transfer pipe rupture accidents: according to the state of a two-loop system in the event of an SGTR (serving gateway controller) accident, the method can judge that the SGTR superimposed two-loop fault occurs, and comprises the following steps:
judging whether a two-loop pipeline break accident is superposed or not according to the pressure of the damaged steam generator;
judging whether a safety valve or a release valve is superposed or not according to the state of a safety valve or an atmospheric release valve of the damaged steam generator;
judging whether the accident that the damaged SG cannot be isolated is superposed or not according to the state of the damaged SG isolation valve;
and judging whether the secondary side heat trap loss accident is superposed or not according to the damaged SG feed water flow and the like.
To steam generator heat-transfer pipe rupture accident stack secondary side trouble, operator's operating point includes: the primary loop system is cooled and depressurized as soon as possible to reduce loss of coolant, safety injection flow is reduced in time, safety injection is stopped in time to reduce possibility of overflow of a steam generator, cold shutdown is realized at the fastest allowable speed or the fastest speed which can be achieved by the power plant according to actual states of the power plant, the primary loop system is depressurized to stop leakage of radioactive substances, a normal surplus discharge system is put into use, and finally the reactor device is brought into a cold shutdown state.
Through the steam generator heat transfer tube rupture accident judgment and operation key points, an operator can adopt the most appropriate accident handling strategy to handle the accident according to the full-range steam generator heat transfer tube rupture accident handling strategy designed in the embodiment 1, and the radioactivity release after the accident is reduced.
The relevant equipment involved in a full-scale steam generator heat transfer tube rupture management strategy is primarily an operator interface. The operator interface is a control interface that provides a signal indication of "nuclear measurement system status", "steam generator level/pressure", "pressurizer level/pressure", "core outlet subcooling", "containment pressure and radioactivity dose rate", "main pump operating status", "steam generator radioactivity monitoring", and pressurizer safety valve and steam generator safety valve and atmospheric relief valve status, as well as "safety injection system", "containment isolation system", "containment spray system", "chemical and volume control system", "steam generator isolation system", "steam generator auxiliary water supply system", "steam generator atmospheric vent system", "pressurizer pressure control system", "steam generator blowdown system", and the like.
The embodiment 1 is an accident coping strategy after a full-range steam generator heat transfer tube rupture accident (including a steam generator heat transfer tube rupture accident and a primary loop or a secondary loop accident superimposed on the steam generator heat transfer tube rupture accident) of a pressurized water reactor nuclear power plant (station). The main purpose of this example 1 is to determine the specific steam generator heat transfer tube rupture accident category according to the actual response of the power plant after the steam generator heat transfer tube rupture accident occurs, and to reduce the radioactive emission of the damaged steam generator by setting a reasonable accident countermeasure strategy. The strategy expands the coping range of the steam generator heat transfer pipe rupture accident handling strategy and optimizes the steam generator heat transfer pipe rupture accident handling strategy.
Example 2
Example 2 is a full-range steam generator heat transfer tube rupture accident handling strategy, and the operation flow is shown in fig. 1, and specifically as follows:
firstly, according to accident characteristics such as monitoring of leakage quantity N-16 of a steam generator, total gamma activity of steam, gamma activity of steam extracted by a condenser, abnormal gamma activity of sewage discharged by the steam generator or uncontrollable rising of water level of a narrow range of the steam generator and the like, confirming that a steam generator heat transfer pipe is broken;
secondly, after confirming that the accident of the heat transfer pipe of the steam generator occurs, judging the actually occurring accident type of the rupture of the heat transfer pipe of the steam generator according to the state of the power plant:
1. general steam generator heat transfer tube rupture accident:
the state parameters of the one-loop system and the two-loop system are normal.
2. Steam generator heat-transfer pipe breaks stack a return circuit loss of coolant accident or single steam generator many heat-transfer pipe accidents of breaking:
the safety valve of the voltage stabilizer is blocked; after the safety valve of the voltage stabilizer is closed, the pressure of a loop continuously decreases; the pressure of a primary circuit is low; the supercooling degree of the reactor core outlet is continuously reduced; the safety injection can not be terminated; restarting the safety injection after the safety injection is terminated; the safety injection box cannot be isolated; and reaching the manual starting criterion of the folded page safety notes.
3. The pressure of a primary loop is out of control after the heat transfer tubes of the steam generator are broken and superposed:
the pressurizer main spray is not available, and the pressurizer auxiliary spray is not available, and the pressurizer safety valve is not available.
4. The steam generator heat transfer pipe breaks and overlaps two circuit faults:
a damaged SG cannot be isolated from a perfect SG; the broken SG safety valve or the atmospheric relief valve is blocked; the pressure of damaged SG is low; the secondary side heat sink is lost.
And finally, selecting different accident coping strategies according to the actually occurring steam generator heat transfer pipe rupture accident types:
1. general steam generator heat transfer tube rupture accident:
identifying and isolating all damaged steam generators; cooling the loop system by using a complete steam generator; reducing the pressure of a primary circuit to recover the water charge of the primary circuit; stopping safety injection to stop leakage of the primary side to the secondary side; determining an optimal post-SGTR accident cooling method according to the state of the power plant; cooling the loop system to enable the loop system to reach the access temperature of the waste heat discharge system; reducing the secondary pressure of a loop and a damaged steam generator synchronously to reach the access pressure of a waste heat discharge system; and the normal waste heat discharge system is utilized to enable the reactor to reach a cold shutdown state.
2. Steam generator heat-transfer pipe breaks stack a return circuit loss of coolant accident or single steam generator many heat-transfer pipe accidents of breaking:
identifying and isolating all damaged steam generators; the pressure of a primary circuit system is reduced as soon as possible so as to reduce the loss of the primary circuit coolant; reducing the flow of the safety injection in time and stopping the safety injection in time so as to reduce the possibility of overflowing of the steam generator; the method includes the steps of selecting a cold shutdown condition to be achieved at the fastest allowable rate or the maximum allowable cooling rate achievable by the power plant according to actual conditions of the power plant, reducing the pressure of a primary loop to the atmospheric pressure to stop leakage of radioactive materials, and if the water level of a damaged steam generator is too high, achieving cold shutdown by an operator at the maximum allowable cooling rate achievable by the power plant in order to reduce radioactive release to the environment.
3. The pressure of a primary loop is out of control after the heat transfer tubes of the steam generator are broken and superposed:
identifying and isolating all damaged steam generators; cooling the loop system by using a complete steam generator; recovering the primary circuit pressure control, and if the primary circuit pressure control can be recovered, executing a general steam generator heat transfer pipe rupture accident coping strategy; if the loop pressure control cannot be recovered, continuing to execute a subsequent operation strategy; stopping safety injection to stop leakage of the primary side to the secondary side; continuously cooling the loop system to enable the loop system to reach the access temperature of the waste heat discharge system; reducing the secondary pressure of a loop and a damaged steam generator to reach the access pressure of a waste heat discharge system; and the normal waste heat discharge system is utilized to enable the reactor to reach a cold shutdown state.
4. The steam generator heat transfer pipe breaks and overlaps two circuit faults:
identifying and isolating all damaged steam generators; the pressure of a primary circuit system is reduced as soon as possible so as to reduce the loss of the primary circuit coolant; reducing the flow of the safety injection in time and stopping the safety injection in time so as to reduce the possibility of overflowing of the steam generator; according to the actual state of the power plant, selecting to reach a cold shutdown state at an allowable fastest speed or a maximum allowable cooling rate which can be reached by the power plant, and simultaneously reducing the pressure of a primary circuit to the atmospheric pressure to stop the leakage of the radioactive substances; if the water level of the built-in refueling water tank is too low or the water level of the damaged steam generator is too high, in order to preserve emergency core cooling water and reduce radioactive release to the environment, an operator should realize cold shutdown at the maximum cooling rate achievable by the power plant.
The full-range steam generator heat transfer tube rupture accident coping strategy of the embodiment 2 can effectively cope with the steam generator heat transfer tube rupture type accidents: and stopping the accident of primary circuit leakage before the reactor reaches a cold shutdown state by adopting a safety injection stopping strategy. The method mainly comprises the steps that one or more steam generator heat transfer pipe rupture accidents are avoided, the steam generator heat transfer pipe rupture accident is superposed with a primary loop pressure out of control, or the steam generator heat transfer pipe rupture accident is superposed with a good steam generator secondary loop fault; and a primary circuit continuously leaks, and the accident of primary circuit leakage is limited by reducing the safety injection flow. The method mainly comprises the steps of overlapping the rupture accident of a heat transfer pipe of the steam generator with the damage accident of a loop water loss or a secondary loop breach of the steam generator.
Example 3
This embodiment 3 is a PWR nuclear power plant full range SGTR accident handling system, includes:
a monitoring module: the system is used for monitoring the states of all equipment in the pressurized water reactor nuclear power plant;
a judging module: the system is used for judging whether a steam generator heat transfer pipe rupture accident occurs or not according to the states of all equipment in the pressurized water reactor nuclear power plant, judging whether the steam generator heat transfer pipe rupture accident has a superposition accident or not and judging the type of the steam generator heat transfer pipe rupture accident; the types of accidents of the steam generator heat transfer pipe breakage comprise: the general type steam generator heat transfer pipe rupture accident, the steam generator heat transfer pipe rupture accident superimposed primary loop water loss accident, the steam generator heat transfer pipe rupture accident superimposed primary loop pressure out of control, the steam generator heat transfer pipe rupture accident superimposed secondary side fault and the like.
A processing module: the accident handling method comprises the steps of establishing a corresponding accident handling process according to the type of accidents such as breakage of a heat transfer pipe of the steam generator;
a display module: the method is used for displaying the type of the steam generator heat transfer pipe rupture accident and the accident handling process corresponding to the type of the accident.
The full-range SGTR accident handling system of the pressurized water reactor nuclear power plant in the embodiment 3 can be used for handling full-range steam generator heat transfer pipe breakage accidents, particularly steam generator heat transfer pipe breakage accident superposition accidents, and fills the blank in the prior art.
The above-mentioned embodiments are intended to illustrate the objects, technical solutions and advantages of the present invention in further detail, and it should be understood that the above-mentioned embodiments are merely exemplary embodiments of the present invention, and are not intended to limit the scope of the present invention, and any modifications, equivalent substitutions, improvements and the like made within the spirit and principle of the present invention should be included in the scope of the present invention.
Claims (10)
1. A full-range SGTR accident handling method for a pressurized water reactor nuclear power plant is characterized by comprising the following steps:
s1: judging whether a steam generator heat transfer pipe rupture accident occurs or not according to the narrow-range water level of the steam generator and the two-loop radioactive signal;
s2: judging the type of the steam generator heat transfer tube rupture accident after the steam generator heat transfer tube rupture accident occurs, wherein the type of the steam generator heat transfer tube rupture accident comprises the following steps: a first type of accident, a second type of accident, a third type of accident, and a fourth type of accident;
the first type of accident is a general type steam generator heat transfer pipe rupture accident, the second type of accident is a steam generator heat transfer pipe rupture accident superimposed primary loop water loss accident, the third type of accident is a steam generator heat transfer pipe rupture accident superimposed primary loop pressure runaway, and the fourth type of accident is a steam generator heat transfer pipe rupture accident superimposed secondary side fault;
s3: and according to the type of the steam generator heat transfer pipe breakage accident, using a corresponding accident handling step.
2. The SGTR full-scope accident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein the step S2 includes:
judging whether the steam generator heat transfer pipe breakage accident is a first accident or not according to the state of the secondary side of the intact steam generator;
judging whether the steam generator heat transfer pipe breakage accident is a second accident or not according to the state of a loop system;
judging whether the steam generator heat transfer pipe breakage accident is a third accident or not according to the condition whether main spray, auxiliary spray and a safety valve of the pressure stabilizer are unavailable or not;
and judging whether the steam generator heat transfer pipe breakage accident is a fourth accident or not according to the state of the two-circuit system.
3. The method for handling the SGTR event in the full range of a pressurized water reactor nuclear power plant according to claim 2, wherein the second type of event comprises a steam generator heat transfer tube rupture event superimposed stabilizer safety valve stuck-open event, a single steam generator multi-heat transfer tube rupture event and a steam generator heat transfer tube rupture event superimposed primary loop breach event.
4. The method for handling the SGTR event throughout the PWR nuclear power plant according to claim 3,
judging whether the safety valve of the superposed voltage stabilizer is blocked according to the state of the safety valve of the voltage stabilizer;
and judging whether a primary loop break accident is superposed or a plurality of heat transfer tubes of a single steam generator break accident occurs according to whether the primary loop pressure is too low, the supercooling degree of the reactor core outlet is continuously reduced, the safety injection pump cannot be stopped, the safety injection pump needs to be restarted after being stopped or the safety injection tank cannot be isolated.
5. The method for handling the SGTR event throughout the PWR nuclear power plant according to claim 2, wherein the secondary side fault comprises: the accident of two-loop pipeline breakage, the accident of steam generator safety valve or release valve blockage, the accident of damaged steam generator unable isolation and the accident of secondary side heat trap loss.
6. The method for handling the SGTR event throughout the PWR nuclear power plant according to claim 5;
judging whether the secondary side fault is a two-loop pipeline break accident or not according to the pressure of the damaged steam generator;
judging whether the secondary side fault is a safety valve or release valve stuck-open accident or not according to the state of a safety valve or an atmospheric release valve of the damaged steam generator;
judging whether the secondary side fault is a failure of isolating the damaged steam generator or not according to the state of an isolating valve of the damaged steam generator;
and judging whether the secondary side fault is a secondary side heat trap loss accident or not according to the water supply flow of the damaged steam generator.
7. The SGTR full-scope accident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein the step S3 includes:
when a first type accident occurs, the accident handling steps are as follows: identifying and isolating the damaged steam generator, cooling a primary loop system to establish supercooling degree, reducing the pressure of the primary loop to recover the water content of the primary loop system, stopping safety injection to stop leakage from the primary side to the secondary side, and synchronously reducing the pressure of the primary loop system and the primary and secondary sides of the damaged steam generator to a residual discharge access working condition, and finally maintaining the nuclear power plant in a cold shutdown working condition;
when a second type of accident occurs, the accident handling steps are as follows: the primary loop system cools and reduces the pressure as soon as possible to reduce the loss of the coolant, reduces the flow of safety injection in time and stops the safety injection in time to reduce the possibility of overflow of a steam generator, realizes cold shutdown and primary loop system pressure reduction to stop leakage of radioactive substances according to the actual state of the power plant at the allowed fastest speed or the fastest speed which can be achieved by the power plant, puts into a normal waste heat discharge system, and finally brings the reactor device into a cold shutdown state;
when a third type of accident occurs, the accident handling steps are as follows: recovering pressure control of the voltage stabilizer, stopping safety injection to stop leakage of the primary side and the secondary side, preparing and starting to cool a loop system, putting into a normal waste heat discharge system, and finally bringing the reactor into a cold shutdown state;
when a fourth type of accident occurs, the accident handling steps are as follows: the primary loop system cools and reduces the pressure as soon as possible to reduce the loss of the coolant, reduces the flow of safety injection in time and stops the safety injection in time to reduce the possibility of overflow of the steam generator, realizes cold shutdown according to the actual state of the power plant at the allowed fastest speed or the fastest speed which can be achieved by the power plant, reduces the pressure of the primary loop system to stop leakage of radioactive substances, puts into a normal waste heat discharge system, and finally brings the reactor device into a cold shutdown state.
8. The method for handling the SGTR event at a full scale of a pressurized water reactor nuclear power plant according to claim 1, wherein the first type of event comprises: a single steam generator tube burst event, two steam generator tube burst events, and multiple steam generator tube burst events.
9. The method for the full-scale SGTR incident handling of a pressurized water reactor nuclear power plant according to claim 1, wherein the two-loop activity signal includes: monitoring the leakage quantity N-16 of the steam generator, the total gamma activity of the steam, the gamma activity of the air extracted by the condenser and the gamma activity of the sewage discharged by the steam generator.
10. A full-scale SGTR accident handling system for a pressurized water reactor nuclear power plant, comprising:
a monitoring module: the system is used for monitoring the states of all equipment in the pressurized water reactor nuclear power plant;
a judging module: the system is used for judging whether a steam generator heat transfer pipe rupture accident occurs or not according to the states of all equipment in the pressurized water reactor nuclear power plant, judging whether the steam generator heat transfer pipe rupture accident has a superposition accident or not and judging the type of the steam generator heat transfer pipe rupture accident;
a processing module: the accident handling method comprises the steps of establishing a corresponding accident handling process according to the type of the steam generator heat transfer pipe breakage accident;
a display module: and the accident type and the processing flow thereof are displayed.
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