CN111540487A - Cooling treatment method for reactor after steam generator heat transfer pipe failure accident - Google Patents
Cooling treatment method for reactor after steam generator heat transfer pipe failure accident Download PDFInfo
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- CN111540487A CN111540487A CN202010362381.3A CN202010362381A CN111540487A CN 111540487 A CN111540487 A CN 111540487A CN 202010362381 A CN202010362381 A CN 202010362381A CN 111540487 A CN111540487 A CN 111540487A
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
- G21C15/18—Emergency cooling arrangements; Removing shut-down heat
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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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Abstract
The scheme discloses a cooling treatment method for a reactor after a steam generator heat transfer tube breakage accident, which is used for realizing the cooling and depressurization of the reactor to a cold shutdown state after the steam generator heat transfer tube breakage accident happens to a pressurized water reactor, and comprises the following steps in sequence: s1, stopping fluid leakage between the first loop and the second loop after the accident; s2, cooling a loop by using a steam generator which normally works, and reducing the pressure of the loop by using a voltage stabilizer on the loop; s3, when the pressure of the primary loop is lower than the pressure of the secondary loop, the working medium on the secondary side of the damaged steam generator is injected back into the primary side of the steam generator, so that the pressure and the temperature of the primary loop are further reduced in pressure and temperature; and S4, connecting a waste heat discharge system to enable the reactor to reach a cold shutdown working condition. By adopting the scheme, the temperature and the pressure are reduced after the SGTR accident is realized, and meanwhile, the radioactivity release can be stopped or reduced to the minimum.
Description
Technical Field
The invention relates to the technical field of pressurized water reactor accident handling, in particular to a cooling treatment method for a reactor after a steam generator heat transfer pipe breaks an accident.
Background
After a steam generator heat transfer tube is broken (SGTR, heat transfer tube breakage accident) in a pressurized water reactor nuclear power plant, a primary loop coolant leaks to the broken Steam Generator (SG) due to the fact that the pressure of a primary loop is far larger than the pressure of the broken Steam Generator (SG) (secondary side pressure), and loss of the primary loop coolant is caused, and meanwhile, the radioactivity of a secondary loop system is increased.
Under the accident, the existing accident handling method is as follows: the operator first identifies and isolates the broken SG limit radioactivity release, after ensuring primary circuit supercooling degree and water charge, stops the safety injection system to stop the primary side leakage to the secondary side. After the safety injection is stopped, the nuclear power plant is stabilized in a hot shutdown working condition, and still needs to be cooled and depressurized to a cold shutdown state to ensure that no further radioactivity is released, and repair the damaged SG heat transfer pipe. Since the broken SG inhibits the step-down process of the loop as in the second regulator, the step-down and step-down operations after the SGTR accident are complicated.
The existing nuclear power plant usually adopts a steam generator pollution discharge or steam exhaust mode to carry out subsequent temperature reduction and pressure reduction, but the two modes have the potential defect of expanded radioactive release range.
Disclosure of Invention
Aiming at the problem that the steam generator pollution discharge or steam exhaust mode is generally adopted for subsequent cooling and pressure reduction in the existing nuclear power plant, but both the two modes have the potential defect of expanded radioactive release range, the scheme provides a reactor cooling treatment method after a steam generator heat transfer pipe breakage accident.
The technical means of the scheme is as follows, the method for cooling the reactor after the steam generator heat transfer tube breakage accident is used for cooling and depressurizing the reactor to a cold shutdown state after the steam generator heat transfer tube breakage accident occurs to the pressurized water reactor, and the method comprises the following steps which are carried out in sequence:
s1, stopping fluid leakage between the first loop and the second loop after the accident;
s2, cooling a loop by using a steam generator which normally works, and reducing the pressure of the loop by using a voltage stabilizer on the loop;
s3, when the pressure of the primary loop is lower than the pressure of the secondary loop, the working medium on the secondary side of the damaged steam generator is injected back into the primary side of the steam generator, so that the pressure and the temperature of the primary loop are further reduced in pressure and temperature;
and S4, connecting a waste heat discharge system to enable the reactor to reach a cold shutdown working condition.
The invention aims to provide an accident countermeasure aiming at a pressurized water reactor nuclear power plant (station), and provides a method for realizing the cooling and depressurization of a reactor system in a back injection type cooling mode under the working condition of an SGTR accident, and simultaneously reducing the radioactivity release to the minimum or avoiding the radioactivity release, so that a primary circuit system is smoothly cooled to the working condition of cold shutdown, further no radioactivity release is ensured, and the condition of repairing a damaged SG heat transfer pipe is achieved.
The basic principle of the technical scheme is as follows: when the pressurized water reactor nuclear power plant (station) generates SGTR, the reactor system is cooled and depressurized in a back injection type cooling mode. In step S1, after the leakage of the first loop to the second loop through the broken heat transfer tube is blocked, step S2 is performed, the intact steam generator is used to cool the primary loop (primary loop) system at a certain rate, and under the condition of ensuring enough supercooling degree of the primary loop medium at the core outlet, if the pressure stabilizer is used to normally spray the primary loop system, the pressure of the primary loop system is continuously reduced, and if the primary loop pressure is lower than the secondary pressure of the damaged steam generator, the SG secondary side water charge is back-injected to the primary loop system through the breach under the action of pressure difference, so as to realize the temperature and pressure reduction of the primary and secondary loop systems, and simultaneously, the coolant of the primary loop system with radioactivity is prevented from being discharged through the damaged steam generator, thereby avoiding or minimizing the radioactivity release. When the temperature in a loop is reduced to below 170 ℃ and the pressure in the loop is reduced to below 4Mpa, the existing waste heat discharge system is utilized to cool and reduce the temperature of the reactor to a cold shutdown working condition.
The further technical scheme is as follows:
to avoid that in step S3, the fluid fed into the primary circuit causes the boron concentration in the primary circuit to decrease, which results in the reactor not having enough shutdown depth, the method is set as follows: before the working medium on the secondary side of the damaged steam generator is back injected to the primary side of the steam generator, the method further comprises a boron concentration detection step, wherein the boron concentration detection step is as follows: and detecting the boron concentration of the primary loop of the reactor and the secondary side of the damaged steam generator. The detection values obtained in the above boron concentration detection step are used as guidance such as: in step S3, the amount of fluid in the circuit is supplemented or an amount of boron is further added to the circuit.
As a technical solution that can implement step S3 based on the existing system and equipment, the following are provided: in step S3, a core-cooling medium is replenished into the primary circuit via the secondary side of the damaged steam generator;
in order to ensure the cooling effect on a loop and the shutdown depth of a reactor, the method is set as follows: the boron concentration value obtained in the boron concentration detection step is used for calculating the amount of boron added into a loop. As a person skilled in the art, the secondary side of the steam generator is supplemented with a loop, and the secondary side and the primary side of the steam generator are communicated through the break.
More completely, as a technical solution based on the existing system and device, the step S3 can be implemented, and the technical solution is set as: the cooling medium for the reactor core is supplemented into a loop through the secondary side of the damaged steam generator and is realized through the pressure difference between the primary side and the secondary side of the damaged steam generator.
As a method for not only achieving the leakage termination in the step S1 but also preventing radioactive substances from contaminating the inlet side and the outlet side of the steam generator as much as possible, it is configured such that: the specific implementation method of the step S1 is as follows: and closing valves on a water inlet and an air outlet of the damaged steam generator, and realizing pressure equalization of the primary side and the secondary side of the heat transfer pipe through the broken heat transfer pipe. By adopting the scheme, the communication paths of the damaged steam generator and the water inlet side and the air outlet side are cut off by closing the valves, and if the communication paths are compared with a loop pipeline between the damaged steam generator and the pressure vessel, the radioactive substances entering the secondary side of the damaged steam generator can be prevented from further diffusing as much as possible.
Considering the thermal stress loaded on the reactor system in the cooling process, for protecting the reactor system, the following settings are set: in step S2, the core primary cold leg cool-down rate is maintained at less than 55 ℃/h while cooling the primary loop with the steam generator in normal operation.
The liquid level within the broken steam generator is monitored prior to performing step S3. This scheme aims at such as: if the broken SG water level is below 5% narrow range, there is a possibility of overlap occurrence of main steam pipe or main water supply pipe rupture, so it is set to include level monitoring to guide the operator to the relevant regulations for treatment.
As a specific loop voltage reduction mode, the following steps are set: the specific implementation mode of utilizing the voltage stabilizer on the loop to reduce the pressure of the loop is as follows: and the voltage reduction is realized by normal spraying of the voltage stabilizer. More specifically, the supercooling degree of the reactor core outlet is kept to be more than 27 ℃ in the depressurization process so as to avoid the integral boiling of the reactor core coolant.
To maintain the liquid level in the circuit, the arrangement is: before step S3 is performed, the water level in the pressurizer is maintained.
As a specific liquid level maintaining specific embodiment, the liquid level maintaining device is provided with the following steps: the specific implementation manner of the maintenance control is as follows: the method is realized by controlling the charging and discharging flow of the voltage stabilizer.
The invention has the following beneficial effects:
the invention aims to provide an accident countermeasure aiming at a pressurized water reactor nuclear power plant (station), and provides a method for realizing the cooling and depressurization of a reactor system in a back injection type cooling mode under the working condition of an SGTR accident, and simultaneously reducing the radioactivity release to the minimum or avoiding the radioactivity release, so that a primary circuit system is smoothly cooled to the working condition of cold shutdown, further no radioactivity release is ensured, and the condition of repairing a damaged SG heat transfer pipe is achieved.
The basic principle of the technical scheme is as follows: when the pressurized water reactor nuclear power plant (station) generates SGTR, the reactor system is cooled and depressurized in a back injection type cooling mode. In step S1, after the leakage of the first loop to the second loop through the broken heat transfer tube is blocked, step S2 is performed, the intact steam generator is used to cool the primary loop (primary loop) system at a certain rate, and under the condition of ensuring enough supercooling degree of the primary loop medium at the core outlet, if the pressure stabilizer is used to normally spray the primary loop system, the pressure of the primary loop system is continuously reduced, and if the primary loop pressure is lower than the secondary pressure of the damaged steam generator, the SG secondary side water charge is back-injected to the primary loop system through the breach under the action of pressure difference, so as to realize the temperature and pressure reduction of the primary and secondary loop systems, and simultaneously, the coolant of the primary loop system with radioactivity is prevented from being discharged through the damaged steam generator, thereby avoiding or minimizing the radioactivity release. When the temperature in a loop is reduced to below 170 ℃ and the pressure in the loop is reduced to below 4Mpa, the existing waste heat discharge system is utilized to cool and reduce the temperature of the reactor to a cold shutdown working condition.
Detailed Description
The present invention will be described in further detail with reference to examples, but the structure of the present invention is not limited to the following examples.
Example 1:
a cooling treatment method for a reactor after a steam generator heat transfer tube breakage accident is used for realizing the cooling and depressurization of the reactor to a cold shutdown state after the steam generator heat transfer tube breakage accident happens to a pressurized water reactor, and the method comprises the following steps which are carried out in sequence:
s1, stopping fluid leakage between the first loop and the second loop after the accident;
s2, cooling a loop by using a steam generator which normally works, and reducing the pressure of the loop by using a voltage stabilizer on the loop;
s3, when the pressure of the primary loop is lower than the pressure of the secondary loop, the working medium on the secondary side of the damaged steam generator is injected back into the primary side of the steam generator, so that the pressure and the temperature of the primary loop are further reduced in pressure and temperature;
and S4, connecting a waste heat discharge system to enable the reactor to reach a cold shutdown working condition.
The invention aims to provide an accident countermeasure aiming at a pressurized water reactor nuclear power plant (station), and provides a method for realizing the cooling and depressurization of a reactor system in a back injection type cooling mode under the working condition of an SGTR accident, and simultaneously reducing the radioactivity release to the minimum or avoiding the radioactivity release, so that a primary circuit system is smoothly cooled to the working condition of cold shutdown, further no radioactivity release is ensured, and the condition of repairing a damaged SG heat transfer pipe is achieved.
The basic principle of the technical scheme is as follows: when the pressurized water reactor nuclear power plant (station) generates SGTR, the reactor system is cooled and depressurized in a back injection type cooling mode. In step S1, after the leakage of the first loop to the second loop through the broken heat transfer tube is blocked, step S2 is performed, the intact steam generator is used to cool the primary loop (primary loop) system at a certain rate, and under the condition of ensuring enough supercooling degree of the primary loop medium at the core outlet, if the pressure stabilizer is used to normally spray the primary loop system, the pressure of the primary loop system is continuously reduced, and if the primary loop pressure is lower than the secondary pressure of the damaged steam generator, the SG secondary side water charge is back-injected to the primary loop system through the breach under the action of pressure difference, so as to realize the temperature and pressure reduction of the primary and secondary loop systems, and simultaneously, the coolant of the primary loop system with radioactivity is prevented from being discharged through the damaged steam generator, thereby avoiding or minimizing the radioactivity release. When the temperature in a loop is reduced to below 170 ℃ and the pressure in the loop is reduced to below 4Mpa, the existing waste heat discharge system is utilized to cool and reduce the temperature of the reactor to a cold shutdown working condition.
Example 2:
this example is further defined on the basis of example 1:
to avoid that in step S3, the fluid fed into the primary circuit causes the boron concentration in the primary circuit to decrease, which results in the reactor not having enough shutdown depth, the method is set as follows: before the working medium on the secondary side of the damaged steam generator is back injected to the primary side of the steam generator, the method further comprises a boron concentration detection step, wherein the boron concentration detection step is as follows: and detecting the boron concentration of the primary loop of the reactor and the secondary side of the damaged steam generator. The detection values obtained in the above boron concentration detection step are used as guidance such as: in step S3, the amount of fluid in the circuit is supplemented or an amount of boron is further added to the circuit.
As a technical solution that can implement step S3 based on the existing system and equipment, the following are provided: in step S3, a core-cooling medium is replenished into the primary circuit via the secondary side of the damaged steam generator;
in order to ensure the cooling effect on a loop and the shutdown depth of a reactor, the method is set as follows: the boron concentration value obtained in the boron concentration detection step is used for calculating the amount of boron added into a loop. As a person skilled in the art, the secondary side of the steam generator is supplemented with a loop, and the secondary side and the primary side of the steam generator are communicated through the break.
More completely, as a technical solution based on the existing system and device, the step S3 can be implemented, and the technical solution is set as: the cooling medium for the reactor core is supplemented into a loop through the secondary side of the damaged steam generator and is realized through the pressure difference between the primary side and the secondary side of the damaged steam generator.
As a method for not only achieving the leakage termination in the step S1 but also preventing radioactive substances from contaminating the inlet side and the outlet side of the steam generator as much as possible, it is configured such that: the specific implementation method of the step S1 is as follows: and closing valves on a water inlet and an air outlet of the damaged steam generator, and realizing pressure equalization of the primary side and the secondary side of the heat transfer pipe through the broken heat transfer pipe. By adopting the scheme, the communication paths of the damaged steam generator and the water inlet side and the air outlet side are cut off by closing the valves, and if the communication paths are compared with a loop pipeline between the damaged steam generator and the pressure vessel, the radioactive substances entering the secondary side of the damaged steam generator can be prevented from further diffusing as much as possible.
Considering the thermal stress loaded on the reactor system in the cooling process, for protecting the reactor system, the following settings are set: in step S2, the core primary cold leg cool-down rate is maintained at less than 55 ℃/h while cooling the primary loop with the steam generator in normal operation.
The liquid level within the broken steam generator is monitored prior to performing step S3.
As a specific loop voltage reduction mode, the following steps are set: the specific implementation mode of utilizing the voltage stabilizer on the loop to reduce the pressure of the loop is as follows: and the voltage reduction is realized by normal spraying of the voltage stabilizer.
To maintain the liquid level in the circuit, the arrangement is: before step S3 is performed, the water level in the pressurizer is maintained.
As a specific liquid level maintaining specific embodiment, the liquid level maintaining device is provided with the following steps: the specific implementation manner of the maintenance control is as follows: the method is realized by controlling the charging and discharging flow of the voltage stabilizer.
The foregoing is a more detailed description of the present invention in connection with specific preferred embodiments thereof, and it is not intended that the specific embodiments of the present invention be limited to these descriptions. For those skilled in the art to which the invention pertains, other embodiments that do not depart from the scope of the invention are intended to be encompassed by the scope of the invention.
Claims (10)
1. A cooling treatment method for a reactor after a steam generator heat transfer tube breakage accident is used for realizing the cooling and depressurization of the reactor to a cold shutdown state after the steam generator heat transfer tube breakage accident happens to a pressurized water reactor, and is characterized by comprising the following steps which are carried out in sequence:
s1, stopping fluid leakage between the first loop and the second loop after the accident;
s2, cooling a loop by using a steam generator which normally works, and reducing the pressure of the loop by using a voltage stabilizer on the loop;
s3, when the pressure of the primary loop is lower than the pressure of the secondary loop, the working medium on the secondary side of the damaged steam generator is injected back into the primary side of the steam generator, so that the pressure and the temperature of the primary loop are further reduced in pressure and temperature;
and S4, connecting a waste heat discharge system to enable the reactor to reach a cold shutdown working condition.
2. The method for cooling the reactor after the steam generator heat transfer pipe failure according to claim 1, further comprising a boron concentration detection step before back injecting the working fluid at the secondary side of the failed steam generator to the primary side of the steam generator, wherein the boron concentration detection step comprises: and detecting the boron concentration of the primary loop of the reactor and the secondary side of the damaged steam generator.
3. The method for post-failure reactor cooling of steam generator heat transfer tubes according to claim 2, wherein in step S3, the core cooling medium is fed into the primary circuit via the secondary side of the failed steam generator;
the boron concentration value obtained in the boron concentration detection step is used for calculating the amount of boron added into a loop.
4. The method as claimed in claim 3, wherein the cooling of the reactor core after the steam generator heat transfer tube failure is achieved by a pressure difference between the primary side and the secondary side of the damaged steam generator by adding the cooling medium to the primary circuit through the secondary side of the damaged steam generator.
5. The method for cooling the reactor after the steam generator heat transfer pipe failure according to claim 1, wherein the step S1 is implemented by: and closing valves on a water inlet and an air outlet of the damaged steam generator, and realizing pressure equalization of the primary side and the secondary side of the heat transfer pipe through the broken heat transfer pipe.
6. The method of claim 1, wherein in step S2, the primary core system cold leg cool-down rate is maintained at less than 55 ℃/h during cooling of the primary loop using a steam generator in normal operation.
7. The method for post-accident reactor cooling of steam generator heat transfer tubes according to claim 1, wherein the liquid level in the damaged steam generator is monitored before step S3 is performed.
8. The method for cooling the reactor after the steam generator heat transfer pipe failure according to claim 1, wherein the pressure reduction of the primary circuit pressure by using the pressurizer on the primary circuit is realized by: and the voltage reduction is realized by normal spraying of the voltage stabilizer.
9. The method for post-accident reactor cooling of a steam generator heat transfer tube according to claim 1, wherein the water level in the pressurizer is maintained and controlled before step S3 is performed.
10. The method for cooling the reactor after the steam generator heat transfer pipe failure according to claim 9, wherein the maintaining control is realized by: the method is realized by controlling the charging and discharging flow of the voltage stabilizer.
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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 |
| CN113421662A (en) * | 2021-06-18 | 2021-09-21 | 中国核动力研究设计院 | Natural circulation cooling method under failure of liquid level indication of pressure vessel of nuclear power plant |
| CN113488214A (en) * | 2021-07-22 | 2021-10-08 | 中国核动力研究设计院 | Natural circulation cooling method for steam on upper end enclosure of pressure vessel of nuclear power plant |
| CN113744902A (en) * | 2021-07-22 | 2021-12-03 | 中国核动力研究设计院 | Natural circulation cooling method for preventing upper seal head of pressure vessel from generating steam in nuclear power plant |
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