CN115331858A - Method for processing SGTR (steam generator and turbine control unit) accident of pressurized water reactor nuclear power plant and control system - Google Patents
Method for processing SGTR (steam generator and turbine control unit) accident of pressurized water reactor nuclear power plant and control system Download PDFInfo
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- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 title claims abstract description 124
- 238000000034 method Methods 0.000 title claims abstract description 39
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- 238000001816 cooling Methods 0.000 claims abstract description 63
- 238000004781 supercooling Methods 0.000 claims abstract description 11
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- 238000002955 isolation Methods 0.000 claims description 25
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- 238000012544 monitoring process Methods 0.000 claims description 14
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- 239000010865 sewage Substances 0.000 claims description 5
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- 230000007274 generation of a signal involved in cell-cell signaling Effects 0.000 claims description 3
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- 239000006066 glass batch Substances 0.000 claims description 2
- 239000002918 waste heat Substances 0.000 abstract description 20
- 238000002347 injection Methods 0.000 abstract description 13
- 239000007924 injection Substances 0.000 abstract description 13
- 230000002285 radioactive effect Effects 0.000 abstract description 13
- 238000007599 discharging Methods 0.000 abstract description 6
- 239000002826 coolant Substances 0.000 description 14
- 230000010485 coping Effects 0.000 description 6
- 238000012546 transfer Methods 0.000 description 5
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Abstract
The invention discloses a method for processing an SGTR accident of a pressurized water reactor nuclear power plant and a control system, wherein firstly, an SGTR accident automatic identification signal is set to isolate a PRS (PRS) of a damaged SG loop, so that the radioactive release of a damaged SG secondary side is controlled; then, the primary circuit is continuously cooled and depressurized, so that the primary circuit can reach a sufficient supercooling degree and a sufficient water charge; then, the pressure of the primary side and the pressure of the secondary side of the damaged SG are preliminarily balanced by controlling the water injection flow of the HPMT, so that the state of the nuclear power plant is stabilized, and the leakage of the radioactive cooling machine on the primary side of the damaged SG to the secondary side is stopped; and finally, the primary loop is cooled to a cold shutdown state, so that a set of scientific and reasonable responding method for the SGTR accident is formed, the SGTR accident of the pressurized water reactor nuclear power plant based on the passive emergency reactor core cooling system and the secondary side passive waste heat discharging system can be effectively responded, and the responding range of the SGTR accident treatment of the nuclear power plant is expanded.
Description
Technical Field
The invention relates to the technical field of nuclear power plant accident handling, in particular to a pressurized water reactor nuclear power plant SGTR accident handling method and a control system.
Background
After a steam generator heat transfer tube rupture accident (SGTR) occurs in a pressurized water reactor nuclear power plant, a primary loop radioactive coolant flows into the secondary side of the damaged steam generator (damaged SG) through a heat transfer tube rupture, and the radioactivity of a secondary loop system is increased. If the off-plant power is lost or the bypass fails during an accident, the radioactive coolant will be vented to the atmosphere through a broken SG atmosphere relief or safety valve. Accordingly, the operator must take appropriate recovery measures to reduce the radioactive emissions of the broken SG. The domestic existing SGTR coping strategy is mainly used for coping with SGTR accidents of active nuclear power plants or SGTR accidents of primary side passive nuclear power plants, and SGTR accidents of pressurized water reactor nuclear power plants designed by passive emergency reactor core cooling systems and secondary side passive waste heat removal systems cannot be effectively coped with.
In view of this, the present application is specifically made.
Disclosure of Invention
The technical problem to be solved by the invention is as follows: the existing SGTR coping method is not suitable for SGTR accidents of a pressurized water reactor nuclear power plant. The purpose is to provide a pressurized water reactor nuclear power plant SGTR accident handling method and control, which can stabilize the state of a nuclear power plant, stop the leakage of a damaged steam generator from the primary side to the secondary side, and reduce the temperature and the pressure of the nuclear power plant to a cold shutdown state after the leakage is stopped.
The invention is realized by the following technical scheme:
on the one hand, the method comprises the following steps of,
the invention provides a method for treating SGTR (steam generator and turbine controller) accidents of a pressurized water reactor nuclear power plant, which comprises the following steps of:
setting an SGTR accident automatic identification signal, and isolating a damaged SG loop PRS according to the SGTR accident automatic identification signal;
identifying SGTR accident characteristics, and executing the following steps according to the SGTR accident characteristics:
continuously cooling the primary circuit until the supercooling degree of the primary circuit meets the requirement;
in the cooling process, continuously reducing the pressure of the primary loop until the water charge of the pressure stabilizer of the primary loop is recovered;
after the temperature and the pressure are reduced, stopping injecting water into the primary loop by the HPMT;
and (3) carrying out secondary cooling on the primary loop by adopting a complete SG (glass batch reactor) loop PRS (continuous strand reactor) until the pressurized water reactor nuclear power plant is in a cold shutdown state.
In a further aspect of the present invention,
the SGTR accident automatic identification signal comprises: the steam generator N16 monitors an abnormal signal, a steam total gamma activity abnormal signal, a condenser air pumping gamma activity abnormal signal, a steam generator sewage gamma activity abnormal signal and a steam generator narrow-range water level out-of-control rising signal.
Further, in the above-mentioned case,
before cooling a loop, the method comprises the following steps: it was confirmed that the water supply side and the steam side of the damaged SG were automatically isolated.
Further, in the above-mentioned case,
confirming that the feedwater side and the steam side of the damaged SG have been automatically isolated comprises the steps of:
confirming that the main water supply isolation valve and the auxiliary water supply isolation valve of the damaged SG are automatically isolated;
confirming that the broken SG loop PRS is automatically isolated;
confirming that the main steam isolation valve and the bypass isolation valve are closed;
increasing the setting value of the atmospheric release valve of the damaged SG;
the blowdown line, drain valves and supply valve of the broken SG are closed.
Further, in the above-mentioned case,
the continuous cooling of a circuit comprises the following steps:
continuously cooling the primary loop by adopting intact SG;
if the intact SG is unavailable, continuously cooling the loop system by adopting an intact SG loop PRS;
and if the intact SG and the intact SG loop PRS are unavailable, continuously cooling the loop system by adopting an RPPS system and a passive emergency cooling system.
In a further aspect of the present invention,
the method for continuously cooling the loop by adopting intact SG comprises the following steps:
if only the SGTR accident happens, controlling the intact SG to continuously cool the primary circuit at a normal cooling rate;
and if the SGTR accident and other accidents happen simultaneously, controlling the intact SG to continuously cool the primary circuit at the highest cooling rate.
In a further aspect of the present invention,
the continuous pressure reduction of the loop comprises the following steps: and a voltage stabilizer is adopted to assist spraying or RPPS to continuously reduce the voltage of a loop.
In a further aspect of the present invention,
after stopping the HPMT from filling the loop, the method comprises the following steps: the core decay heat is removed.
Further, in the above-mentioned case,
the method for removing the decay heat of the reactor core comprises the following steps:
controlling the water content of the primary circuit by controlling the flow of a primary circuit water supplementing system;
the pressure of a loop is controlled by the normal spraying of the voltage stabilizer and the electric heater of the voltage stabilizer.
On the other hand, in the case of a system,
the invention provides a SGTR accident handling control system of a pressurized water reactor nuclear power plant, which comprises:
the monitoring module is used for monitoring the water level of the SG and the radioactivity of the secondary side of the SG, and generating corresponding monitoring abnormal data when the water level of the SG is out of control and rises or the radioactivity of the secondary side of the SG is abnormal;
the signal generation module is used for generating an SGTR accident automatic identification signal according to the monitoring abnormal data;
a PRS isolation control module for automatically identifying signal isolation damage SG loop PRS according to the SGTR accident;
the SGTR accident identification module is used for identifying the type of the SGTR accident automatic identification signal to obtain SGTR accident characteristics;
the cooling system control module is used for controlling the cooling system to continuously cool the primary circuit according to the SGTR accident characteristic until the supercooling degree of the primary circuit meets the requirement;
the pressure reduction system control module is used for controlling the pressure reduction system to continuously reduce the pressure of the primary circuit in the temperature reduction process until the water charge of the pressure stabilizer of the primary circuit is recovered;
the HPMT control module is used for controlling the HPMT to stop injecting water into the primary loop after the temperature reduction and the pressure reduction are finished;
and the PRS control module is used for controlling the PRS of the intact SG loop to carry out secondary cooling on the primary loop until pressurized water is in a cold shutdown state for the nuclear power plant.
Compared with the prior art, the invention has the following advantages and beneficial effects: firstly, an SGTR accident automatic identification signal is set to isolate a broken SG loop PRS, and the radioactivity release of a broken SG secondary side is controlled; then, the primary circuit is continuously cooled and depressurized, so that the primary circuit can reach a sufficient supercooling degree and a sufficient water charge; then, the pressure of the primary side and the pressure of the secondary side of the damaged SG are preliminarily balanced by controlling the water injection flow of the HPMT, so that the state of the nuclear power plant is stabilized, and the leakage of the radioactive cooler on the primary side of the damaged SG to the secondary side is stopped; and finally, a primary loop is cooled to a cold shutdown state, so that a set of scientific and reasonable responding method for the SGTR accident is formed, the SGTR accident of the pressurized water reactor nuclear power plant based on the passive emergency reactor core cooling system and the secondary side passive waste heat removal system can be effectively responded, and the responding range of the SGTR accident treatment of the nuclear power plant is expanded.
Drawings
In order to more clearly illustrate the technical solutions of the exemplary embodiments of the present invention, the drawings that are required to be used in the embodiments will be briefly described below, it should be understood that the following drawings only illustrate some embodiments of the present invention and therefore should not be considered as limiting the scope, and that for those skilled in the art, other related drawings can be obtained from these drawings without inventive effort.
Fig. 1 is a schematic diagram of a pressurized water reactor loop configuration provided in embodiment 1 of the present invention;
fig. 2 is a schematic flow chart of a SGTR accident handling method for a pressurized water reactor nuclear power plant according to embodiment 1 of the present invention;
fig. 3 is a schematic structural diagram of an SGTR accident handling control system of a pressurized water reactor nuclear power plant according to embodiment 2 of the present invention.
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
The existing domestic SGTR coping strategy is mainly used for coping with SGTR accidents of active nuclear power plants or SGTR accidents of primary side passive nuclear power plants, and cannot effectively cope with SGTR accidents of pressurized water reactor nuclear power plants (stations) designed by passive emergency reactor core cooling systems and secondary side passive waste heat removal systems.
The passive emergency cooling system is provided with a full-pressure water replenishing tank (HPMT), a safety injection tank, an outer-shell pressure-bearing water tank (RWST), a pressure relief protection system (RPPS) and the like, and has the safety function of providing sufficient water replenishing for a reactor coolant system after an accident; the safety function of the secondary side passive waste heat removal system (PRS) is to automatically put into operation after an accident so as to remove the waste heat of the reactor core.
Fig. 1 shows the structure of a pressurized water reactor circuit. An SGTR event may result in a reduction in the reactor coolant system water inventory, but may not result in significant core damage, but may only result in leakage of primary radioactive coolant to the secondary loop system via a breach in the heat transfer tubes. The SGTR accident can cause the water content of a reactor coolant system to be continuously reduced, and the reactor emergency shutdown, the full-pressure water replenishing tank (HPMT) of the passive emergency cooling system and the secondary side passive waste heat discharging system are triggered to be automatically put into operation. Wherein, the HPMT injects boron-containing water into the reactor coolant system in a water recirculation operation mode to maintain the water content of the reactor system; and a secondary side passive residual heat removal system (PRS) is automatically put into operation to take away the residual heat of the reactor core, and the related valves of the second loop of the damaged SG loop are automatically isolated. Since the PRS system is located on the secondary side, the continuous operation of a broken SG loop PRS reduces the broken SG pressure, resulting in difficult balancing of the broken SG primary and secondary sides.
Therefore, according to the SGTR accident handling method for the pressurized water reactor nuclear power plant provided by the embodiment, firstly, an SGTR accident automatic identification signal is set (the PRS system is started to meet the SGTR accident automatic identification signal), and according to the SGTR accident automatic identification signal, the damaged loop PRS system automatically isolates the relevant valves of the second loop of the damaged SG loop, so that the pressure of the damaged SG gradually rises; the good loop PRS system is then used for loop cooling. As the loop system cools and the water charge decreases, the water level and pressure of the pressurizer decrease, the HPMT injection flow will balance with the break flow and the loop coolant shrinkage, and the loop pressure will approach a balance value. Of course, a pressurizer spray or pressure relief protection system (RPPS) may also be used to manually reduce the reactor coolant system pressure until the SG primary and secondary pressures are balanced and the SGTR breach flow is terminated. The water quantity of the full-pressure water supply tank (HPMT) is enough to ensure the safety of the reactor, a pressure relief protection system (RPPS) cannot be automatically started in the accident process, the reactor core cannot be exposed, the safety injection tank does not need to be put into use, and the secondary pressure balance of the damaged SG can be ensured only by automatically putting into operation a passive emergency cooling system and a secondary side passive waste heat discharge system (PRS). After the state of a nuclear power plant (station) is stable, when the HPMT isolation condition is reached, the HPMT needs to be isolated manually, and a loop water supplement system is adopted to maintain the liquid level of the voltage stabilizer; cooling the loop system by adopting a PRS system or a perfect steam generator (or a normal waste heat discharge system); the pressure of a primary loop can also be reduced by adopting a pressure stabilizer for normal spraying or an RPPS valve until a cold shutdown state is reached.
According to the technical concept, the method for processing the SGTR accident of the pressurized water reactor nuclear power plant comprises two parts of setting an SGTR accident automatic protection signal and formulating an SGTR accident relieving means, and the implementation flow is shown in fig. 1. The method specifically comprises the following steps:
step 1: and setting an SGTR accident automatic identification signal, and isolating the damaged SG loop PRS according to the SGTR accident automatic identification signal.
After a steam generator heat transfer pipe rupture accident occurs, primary side radioactive coolant leaks to a two-loop system through a break, the water level of a voltage stabilizer continuously drops, secondary side radioactivity of the steam generator is abnormal, a main water supply control system of the steam generator automatically adjusts the main water supply flow, and when the main water supply control system is unavailable, the damaged SG water level can be caused to rise out of control. In addition, secondary side pressure of damaged SG is continuously reduced due to operation of a secondary side passive waste heat removal system (PRS), primary and secondary pressures of damaged SG are difficult to balance, and overflow of damaged SG is prevented.
Therefore, an SGTR accident automatic identification signal is set, and the signal is automatically identified through the SGTR accident after the PRS is automatically triggered, such as: the steam generator N16 monitors an abnormal signal, a steam total gamma activity abnormal signal, a condenser air pumping gamma activity abnormal signal, a steam generator sewage gamma activity abnormal signal and a steam generator narrow-range water level out-of-control rising signal, a damaged SG loop PRS is isolated, and the pressure of the damaged SG is prevented from continuously falling.
Step 2: and identifying SGTR accident characteristics, and executing steps 3 to 6 according to the SGTR accident characteristics.
After an SGTR accident occurs in a pressurized water reactor nuclear power plant (station) designed based on a passive emergency cooling system and a secondary side passive waste heat removal system (PRS), typical accident characteristics can occur in the case of a damaged SG, such as: the water level of the steam generator rises in a uncontrolled manner within a narrow range, the monitoring of the steam generator N16 is abnormal, the total gamma activity of steam is abnormal, the gamma activity of sewage discharged by the steam generator is abnormal, and the like.
Therefore, an operator needs to quickly diagnose the SGTR accident according to the steam generator heat transfer pipe rupture accident characteristics, and perform a recovery operation to stabilize the nuclear power plant (station) state, and terminate the leakage from the primary side to the secondary side; when the water level of the damaged SG is recovered, the water supply of the damaged SG is stopped. The operator may also verify the sequence and status of the automatically launched non-security level system to provide defense in depth. Finally, the nuclear power plant (station) is withdrawn to a cold shutdown state by cooling the primary loop.
It should be noted that once the damaged SG is identified, since the secondary side passive waste heat removal system (PRS) is automatically activated to automatically isolate the damaged SG main water supply isolation valve and the damaged SG steam side valve, an operator needs to confirm that the water supply side and the steam side of the damaged SG are automatically isolated to limit radioactive release as a necessary step to terminate the leakage of one circuit to the second circuit. Confirming whether the feed side and the steam side of the damaged SG have been automatically isolated includes the steps of:
step 2.1: confirming that the main water supply isolation valve and the auxiliary water supply isolation valve of the damaged SG are automatically isolated;
step 2.2: confirming that the broken SG loop PRS is automatically isolated;
step 2.3: confirming that the main steam isolation valve and the bypass isolation valve are closed;
step 2.4: increasing the setting value of the atmospheric release valve of the damaged SG;
step 2.5: the blowdown line, drain valves and supply valve of the broken SG are closed.
If the water supply side and the steam side of the damaged SG are automatically isolated, the PRS system of the damaged SG loop can be isolated according to an automatic isolation signal (the PRS system is started to accord with an SGTR accident automatic identification signal), and the pressure of the damaged SG is prevented from continuously dropping. After the secondary side of the damaged SG is isolated, the pressure of the damaged SG will rise slowly.
And 3, step 3: and continuously cooling the primary circuit until the supercooling degree of the primary circuit meets the requirement. The method comprises the following steps:
step 3.1: continuously cooling the primary loop by adopting intact SG;
step 3.2: if the intact SG is unavailable, continuously cooling the loop system by adopting an intact SG loop PRS;
step 3.3: and if the intact SG and the intact SG loop PRS are unavailable, continuously cooling the loop system by adopting an RPPS system and a passive emergency cooling system.
Following the SGTR event, the primary loop system is automatically started using the good loop PRS (the broken loop PRS has been automatically isolated). If only an SGTR event occurs, then a normal cooling rate may be employed; if a stack accident occurs at the same time, the maximum cooling rate, which is limited by specifications, is required to maintain a primary circuit water charge and the pressure relief protection system (RPPS) need not be triggered. An operator can also reduce the temperature of a loop by using the intact steam generator to provide enough loop supercooling degree for reducing the pressure of the loop system to the damaged SG pressure in the next step, and the method comprises the following two conditions and coping modes: (1) If intact SG is unavailable, cooling by adopting a secondary side passive waste heat discharge system (PRS); under the condition of extremely low probability, the intact SG and the secondary side passive waste heat discharge system are unavailable, and an RPPS system and a passive emergency cooling system can be adopted for cooling.
And 4, step 4: in the cooling process, the pressure of the primary circuit is continuously reduced until the water content of the pressure stabilizer of the primary circuit is recovered. Because the main pump is shut down by the injection signal of the full pressure water supply tank (HPMT), the pressure of the primary circuit can be continuously reduced by using a pressure stabilizer for assisting spraying or RPPS.
During the cooling process of the primary circuit, the injection flow of a full pressure water supplement tank (HPMT) and a primary circuit water supplement system gradually increases the pressure of the primary circuit until the break flow and the total injection flow are balanced. These injection flows must therefore be terminated or controlled to terminate primary-to-secondary leakage, noting that sufficient reactor coolant charge, including sufficient primary loop subcooling and pressurizer water charge, must first be maintained to terminate the HPMT injection before terminating primary-to-secondary leakage. An operator can reduce the pressure of a loop by adopting a voltage stabilizer normal spray or a pressure relief protection system (RPPS); the reactor coolant system pressure can be effectively reduced by means of a secondary side passive waste heat removal system (PRS) alone even if the operator is not active. Leakage from the primary side to the secondary circuit will only stop after the primary circuit pressure and the damaged SG pressure have equalized.
It should be added that the pressure in the primary circuit will continue to drop until the water level in the regulator is restored. Under natural circulation conditions, a vapor cavity may be created by damage to the upper portion of the SG. Since this phenomenon may cause the water level of the pressurizer to rise rapidly, the primary circuit depressurization operation should be terminated when the water level of the pressurizer is high to prevent water from being dense.
And 5: and after the temperature and the pressure are reduced, stopping injecting water into the primary loop by the HPMT to stop leakage of the damaged SG primary side to the secondary side.
Sufficient primary subcooling, secondary side heat traps (two-loop passive waste heat removal (PRS) or intact steam generator) and sufficient primary coolant charge have been established through steps 3 and 4, thus eliminating the need for HPMT injection flow and requiring the operator to terminate the HPMT to terminate primary side to secondary side leakage. After the HPMT injection flow is terminated, the primary side of the damaged SG leaks to the secondary side until the primary circuit pressure and the damaged SG pressure are balanced.
To prevent re-leakage of damaged SGs, operators may remove core decay heat by controlling the primary water makeup system flow, the pressurizer electric heater, and using PRS or a live steam generator (normal waste heat removal system).
And 6: and (4) carrying out secondary cooling on the primary loop by adopting a perfect SG loop PRS until the pressurized water reactor nuclear power plant is in a cold shutdown state.
If the pressure of a primary circuit is balanced with the pressure of the damaged SG secondary side, the leakage of the primary side and the secondary side is stopped, and a primary safety state is achieved. At this time, an operator needs to take a series of operations to cool the reactor to a cold shutdown condition at a controllable cooling rate, including reducing the radioactive contamination of the two loops and restarting the main pump to ensure the uniform coolant temperature and boron concentration, and the operator should select an optimal post-accident cooling mode (discharging steam, discharging sewage and back-injecting to radiate heat to the environment) according to the current power plant state.
In summary, in the method for processing the SGTR accident of the pressurized water reactor nuclear power plant provided by the embodiment, the damaged SG loop PRS is isolated by setting the SGTR accident automatic identification signal, so that the radioactive release of the damaged SG secondary side is controlled; then, the primary circuit is continuously cooled and depressurized, so that the primary circuit can reach a sufficient supercooling degree and a sufficient water charge; then, the pressure of the primary side and the pressure of the secondary side of the damaged SG are preliminarily balanced by controlling the water injection flow of the HPMT, so that the state of the nuclear power plant is stabilized, and the leakage of the radioactive cooling machine on the primary side of the damaged SG to the secondary side is stopped; and finally, a primary loop is cooled to a cold shutdown state, so that a set of scientific and reasonable responding method for the SGTR accident is formed, the SGTR accident of the pressurized water reactor nuclear power plant based on the passive emergency reactor core cooling system and the secondary side passive waste heat removal system can be effectively responded, and the responding range of the SGTR accident treatment of the nuclear power plant is expanded.
Example 2
The embodiment provides a control system corresponding to the SGTR accident handling method of the pressurized water reactor nuclear power plant described in embodiment 1, and is used for realizing digital handling of the SGTR accident handling of the pressurized water reactor nuclear power plant. The SGTR accident handling control system of the pressurized water reactor nuclear power plant is structurally composed as shown in fig. 3, and includes:
the monitoring module is used for monitoring the water level of the SG and the radioactivity of the secondary side of the SG, and generating corresponding monitoring abnormal data when the water level of the SG is out of control and rises or the radioactivity of the secondary side of the SG is abnormal;
the signal generation module is used for generating an SGTR accident automatic identification signal according to the monitoring abnormal data;
a PRS isolation control module for automatically identifying signal isolation damage SG loop PRS according to the SGTR accident;
the SGTR accident identification module is used for identifying the type of the SGTR accident automatic identification signal to obtain SGTR accident characteristics;
the cooling system control module is used for controlling the cooling system to continuously cool the primary circuit according to the SGTR accident characteristics until the supercooling degree of the primary circuit meets the requirement;
the pressure reduction system control module is used for controlling the pressure reduction system to continuously reduce the pressure of the primary circuit in the temperature reduction process until the water content of a pressure stabilizer of the primary circuit is recovered;
the HPMT control module is used for controlling the HPMT to stop injecting water into the primary loop after the temperature reduction and the pressure reduction are finished;
and the PRS control module is used for controlling the complete SG loop PRS to carry out secondary cooling on the primary circuit until the pressurized water is in a cold shutdown state for the nuclear power plant.
It should be added that, the system provides a corresponding control interface for operators to operate the pressurized water to the relevant equipment of the nuclear power plant, and the system comprises:
the system comprises a nuclear measurement system state, a steam generator water level, a steam generator pressure, a voltage stabilizer water level, a voltage stabilizer pressure, a core outlet super-cooling degree, a containment pressure and radioactive dose rate, a main pump operation state, a steam generator secondary side radioactivity monitoring, a main feed water flow, a full pressure water supplementing tank (HPMT) water level, a secondary side passive waste heat discharging system (PRS) state, a primary circuit water supplementing system state, a safety valve of the voltage stabilizer and a safety valve of the steam generator, an atmospheric release valve state, a pressure relief protection system (RPPS) valve state, a spray valve state of the voltage stabilizer, a PRS isolation valve state, a signal indication of the primary steam isolation valve state, a passive emergency cooling system, a containment isolation system, a containment spraying system, a primary circuit water supplementing system, a steam generator isolation system, a steam generator emergency water supply system, a steam generator atmospheric discharge system, a secondary side passive waste heat discharging system, a voltage stabilizer pressure control system, a steam generator blowdown system and other control interfaces.
The above-mentioned embodiments, objects, technical solutions and advantages of the present invention are further described in detail, it should be understood that the above-mentioned embodiments are only examples 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 method for processing SGTR accidents of a pressurized water reactor nuclear power plant is characterized by comprising the following steps:
setting an SGTR accident automatic identification signal, and isolating a damaged SG loop PRS according to the SGTR accident automatic identification signal;
identifying SGTR accident characteristics, and executing the following steps according to the SGTR accident characteristics:
continuously cooling the primary circuit until the supercooling degree of the primary circuit meets the requirement;
in the cooling process, continuously reducing the pressure of the primary loop until the water charge of the pressure stabilizer of the primary loop is recovered;
after the temperature and the pressure are reduced, stopping injecting water into the primary loop by the HPMT;
and (3) carrying out secondary cooling on the primary loop by adopting a complete SG (glass batch reactor) loop PRS (continuous strand reactor) until the pressurized water reactor nuclear power plant is in a cold shutdown state.
2. The SGTR accident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein the SGTR accident automatic identification signal comprises: the steam generator N16 monitors an abnormal signal, a steam total gamma activity abnormal signal, a condenser pumping gamma activity abnormal signal, a steam generator sewage gamma activity abnormal signal and a steam generator narrow-range water level out-of-control rising signal.
3. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein before cooling the primary circuit, the method comprises the following steps: it was confirmed that the water supply side and the steam side of the damaged SG were automatically isolated.
4. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 3, wherein the confirming that the feedwater side and the steam side of the damaged SG have been automatically isolated comprises the steps of:
confirming that the main water supply isolation valve and the auxiliary water supply isolation valve of the damaged SG are automatically isolated;
confirming that the broken SG loop PRS is automatically isolated;
confirming that the main steam isolation valve and the bypass isolation valve are closed;
adjusting the setting value of the atmosphere relief valve of the damaged SG;
the blowdown line, drain valves and supply valve of the broken SG are closed.
5. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein the step of continuously cooling the primary loop comprises the steps of:
continuously cooling the primary loop by adopting intact SG;
if the intact SG is unavailable, continuously cooling the loop system by adopting an intact SG loop PRS;
and if the intact SG and the intact SG loop PRS are unavailable, continuously cooling the primary loop system by adopting an RPPS system and a passive emergency cooling system.
6. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 5, wherein the step of continuously cooling the primary loop with a good SG includes the steps of:
if only the SGTR accident happens, controlling the intact SG to continuously cool the primary circuit at a normal cooling rate;
and if the SGTR accident and other accidents happen simultaneously, controlling the intact SG to continuously cool the primary circuit at the highest cooling rate.
7. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein the step of continuously depressurizing the primary circuit comprises the steps of: and a voltage stabilizer is adopted to assist spraying or RPPS to continuously reduce the voltage of a loop.
8. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 1, wherein after stopping the HPMT from filling the primary circuit, the method comprises the following steps: the core decay heat is removed.
9. The SGTR incident handling method of a pressurized water reactor nuclear power plant according to claim 8, wherein the removing of the core decay heat comprises the steps of:
controlling the water content of the primary loop by controlling the flow of a primary loop water supplementing system;
the pressure of the loop is controlled by the regulator normal spray and the regulator electric heater.
10. A SGTR accident handling control system of a pressurized water reactor nuclear power plant is characterized by comprising:
the monitoring module is used for monitoring the water level of the SG and the radioactivity of the secondary side of the SG, and when the water level of the SG is increased out of control or the radioactivity of the secondary side of the SG is abnormal, corresponding monitoring abnormal data are generated;
the signal generation module is used for generating an SGTR accident automatic identification signal according to the monitoring abnormal data;
the PRS isolation control module is used for automatically identifying signal isolation damage SG loop PRS according to the SGTR accident;
the SGTR accident identification module is used for identifying the type of the SGTR accident automatic identification signal to obtain SGTR accident characteristics;
the cooling system control module is used for controlling the cooling system to continuously cool the primary circuit according to the SGTR accident characteristics until the supercooling degree of the primary circuit meets the requirement;
the pressure reduction system control module is used for controlling the pressure reduction system to continuously reduce the pressure of the primary circuit in the temperature reduction process until the water charge of the pressure stabilizer of the primary circuit is recovered;
the HPMT control module is used for controlling the HPMT to stop injecting water into the primary loop after the temperature reduction and the pressure reduction are finished;
and the PRS control module is used for controlling the PRS of the intact SG loop to carry out secondary cooling on the primary loop until pressurized water is in a cold shutdown state for the nuclear power plant.
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Citations (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104392756A (en) * | 2014-10-08 | 2015-03-04 | 中国科学院合肥物质科学研究院 | Reactor dynamic interlock system and method based on digital instrumentation and control system |
| CN104538068A (en) * | 2013-07-22 | 2015-04-22 | 中国核动力研究设计院 | Method for preventing steam generator from spilling over under heat-transfer tube rupture accident condition |
| CN107068214A (en) * | 2017-05-09 | 2017-08-18 | 中广核研究院有限公司 | Nuclear power plant steam air tapping equipment and secondary circuit pressure method for releasing |
| CN110689973A (en) * | 2019-09-18 | 2020-01-14 | 上海电力大学 | Nuclear power station primary circuit pressure reduction control method under heat transfer pipe fracture accident |
| CN111128414A (en) * | 2019-12-31 | 2020-05-08 | 中国核动力研究设计院 | Active and passive combined safety system and method for nuclear power plant |
| 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 |
| CN111681794A (en) * | 2020-06-19 | 2020-09-18 | 中国核动力研究设计院 | Full-range SGTR accident handling method and system for pressurized water reactor nuclear power plant |
| KR20210098110A (en) * | 2020-01-31 | 2021-08-10 | 한국수력원자력 주식회사 | Analysis Methodologies of Severe Accident Sensitivity Assesment for Steam Generator Tube Rupture in CANDU |
-
2022
- 2022-08-16 CN CN202210980908.8A patent/CN115331858A/en active Pending
Patent Citations (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104538068A (en) * | 2013-07-22 | 2015-04-22 | 中国核动力研究设计院 | Method for preventing steam generator from spilling over under heat-transfer tube rupture accident condition |
| CN104392756A (en) * | 2014-10-08 | 2015-03-04 | 中国科学院合肥物质科学研究院 | Reactor dynamic interlock system and method based on digital instrumentation and control system |
| CN107068214A (en) * | 2017-05-09 | 2017-08-18 | 中广核研究院有限公司 | Nuclear power plant steam air tapping equipment and secondary circuit pressure method for releasing |
| CN110689973A (en) * | 2019-09-18 | 2020-01-14 | 上海电力大学 | Nuclear power station primary circuit pressure reduction control method under heat transfer pipe fracture accident |
| CN111128414A (en) * | 2019-12-31 | 2020-05-08 | 中国核动力研究设计院 | Active and passive combined safety system and method for nuclear power plant |
| KR20210098110A (en) * | 2020-01-31 | 2021-08-10 | 한국수력원자력 주식회사 | Analysis Methodologies of Severe Accident Sensitivity Assesment for Steam Generator Tube Rupture in CANDU |
| 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 |
| CN111681794A (en) * | 2020-06-19 | 2020-09-18 | 中国核动力研究设计院 | Full-range SGTR accident handling method and system for pressurized water reactor nuclear power plant |
Non-Patent Citations (2)
| Title |
|---|
| 徐苗苗 等: ""蒸汽发生器传热管破裂事故缓解措施研究"", 《核安全》, 30 April 2022 (2022-04-30) * |
| 陈伟 等: ""华龙一号快速冷却功能验证试验模拟计算研究"", 《科技视界》, 20 June 2021 (2021-06-20) * |
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