CN110911024A - Sustained release device for steam generator heat transfer tube rupture accident - Google Patents

Sustained release device for steam generator heat transfer tube rupture accident Download PDF

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Publication number
CN110911024A
CN110911024A CN201911150155.2A CN201911150155A CN110911024A CN 110911024 A CN110911024 A CN 110911024A CN 201911150155 A CN201911150155 A CN 201911150155A CN 110911024 A CN110911024 A CN 110911024A
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China
Prior art keywords
valve
safety
containment vessel
release
steam generator
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Pending
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CN201911150155.2A
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Chinese (zh)
Inventor
魏述平
杨洪润
吕焕文
宋丹戎
秦忠
李兰
谭怡
肖锋
朱建平
于红
王军龙
杨舒琦
刘嘉嘉
高希龙
程诗思
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Nuclear Power Institute of China
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Nuclear Power Institute of China
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Priority to CN201911150155.2A priority Critical patent/CN110911024A/en
Publication of CN110911024A publication Critical patent/CN110911024A/en
Pending legal-status Critical Current

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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/04—Safety arrangements
    • G21D3/06—Safety arrangements responsive to faults within the plant
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D1/00—Details of nuclear power plant
    • G21D1/006—Details of nuclear power plant primary side of steam generators
    • 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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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Business, Economics & Management (AREA)
  • Emergency Management (AREA)
  • Structure Of Emergency Protection For Nuclear Reactors (AREA)

Abstract

The invention discloses a slow release device for a steam generator heat transfer pipe rupture accident, which utilizes the slow release device to introduce a large amount of radioactivity generated by an SGTR accident into a containment vessel for control, perfects the integrity of a protective barrier of the containment vessel after the SGTR accident occurs, reduces the radioactive leakage risk caused by the bypass of the containment vessel, reduces the negative influence of the radioactivity on the environment and people, and provides radiation safety guarantee for the related construction of a reactor. The applicant controls the radioactive substances leaked from the steam generator to flow back to the processing equipment in the containment through the pipeline by arranging the three-way valve, so that the radioactive consequences under the conditions of a benchmark accident and a serious accident are reduced, and the safety coefficient of the reactor is improved.

Description

Sustained release device for steam generator heat transfer tube rupture accident
Technical Field
The invention relates to a mounting reference tool of a pressurized water nuclear reactor structure, in particular to a slow release device for a steam generator heat transfer pipe rupture accident.
Background
The occurrence of the fukushima accident leads the public to talk about nuclear color change again, and the international and domestic also put forward higher requirements on radioactivity control, and the actual elimination of a large amount of radioactivity release is no longer a slogan, and the reactor is more required to relate to the proposal of more reasonable and feasible specific measures.
After a Steam Generator heat transfer pipe Rupture accident (SGTR) occurs, the heat transfer pipe in the Steam Generator is ruptured to form a breach, the flow passing through the breach is obviously higher than the leakage rate under normal conditions due to the pressure difference of a first loop and a second loop, the pressure of the second loop is rapidly increased along with the leakage of a primary loop coolant and the tripping of a Steam turbine, a large amount of secondary loop water containing radioactive substances passes through the last safety barrier of a reactor plant in a gaseous state or even in a liquid state and is directly discharged into the environment, and if core melting occurs, a large amount of radioactive substances generated by melting can be discharged into the environment through the breach.
The SGTR accident is an important design basis accident which needs to be evaluated in a safety analysis report of a nuclear power plant, and the radioactive consequence of the SGTR accident cannot meet the requirement of the acceptance criterion of the radioactive consequence in China in some pressurized water reactor nuclear power plants along with the change of the working condition classification of the accident from a limit accident to a rare accident. The SGTR accident has the characteristics of relatively high occurrence frequency and serious radioactive consequences, and the contribution of a common nuclear power plant to a large amount of radioactive release is estimated to be more than 20%. It is therefore of great importance to avoid massive radioactive release in the event of SGTR.
Disclosure of Invention
The invention aims to provide a slow release device for steam generator heat transfer pipe rupture accidents, which reduces the radioactive consequences under the conditions of benchmark accidents and serious accidents and improves the safety and reliability of a reactor.
The invention is realized by the following technical scheme:
the slow release device for the steam generator heat transfer pipe rupture accident comprises a containment vessel of the steam generator and a slow release mechanism, wherein the slow release mechanism introduces radioactive substances released from the containment vessel into the containment vessel after the reactor SGTR accident occurs.
According to the invention, the slow release mechanism is arranged on the containment vessel of the steam generator, and the radioactive substances released from the containment vessel after the SGTR accident of the reactor is generated are reintroduced into the containment vessel through the slow release mechanism, so that the effect of reducing the harmful danger of the radioactive substances of the accident is realized, a radioactive substance collecting device is not required to be specially arranged, and the cost is saved.
Furthermore, a steam generator is arranged in the containment, a steam outlet of the steam generator is connected with a main steam pipeline, the end part of the main steam pipeline penetrates through the containment and then extends outwards, the main steam pipeline is provided with the slow release mechanism, and the slow release mechanism comprises a pressure release device, a return pipe, a safety valve, a three-way control valve, a release valve and an isolation valve. On the basis that the reactor comprises a conventional two-loop system, the radioactive release of the SGTR accident is relieved by reasonably arranging a safety valve, a release valve and an isolation valve on a main steam pipeline and additionally arranging a three-way control valve, a return pipe and pressure relief equipment. Aiming at the problem that when an SGTR accident occurs, a large amount of secondary water containing radioactive substances passes through the last safety barrier of a containment vessel of a reactor in a gaseous state or even in a liquid state and is directly discharged into the environment, if the core is melted, a large amount of radioactive substances generated by melting can also be discharged into the environment through the path, the applicant provides a slow release device for releasing a large amount of radioactive substances caused by the breakage of a heat transfer pipe of a steam generator, namely, the slow release device is used for introducing a large amount of radioactivity generated by the SGTR accident into the containment vessel for control, the integrity of the barrier of the containment vessel after the SGTR occurs is improved, the radioactive leakage risk caused by the bypass of the containment vessel is reduced, the negative influence of the radioactivity on the environment and people is reduced, and the radiation safety guarantee is provided for the related construction of the reactor. The applicant controls the radioactive substances leaked from the steam generator and the main steam pipeline to flow back into the containment through arranging different valve systems and pipelines, so that the radioactive consequences under the conditions of a benchmark accident and a serious accident are reduced.
Further, as a first preferable scheme, the safety valve, the release valve and the isolation valve are all arranged outside the containment vessel, the pressure relief device is arranged in the containment vessel and is connected with one end of the return pipe, the other end of the return pipe is communicated with one end of the safety valve through the three-way control valve, and the other end of the safety valve is communicated with the main steam pipeline; along the fluid output direction, the release valve and the isolation valve are sequentially fixed on the main steam pipeline and are both positioned at the downstream position of the safety valve. The steam and water released from the safety valve are divided into two paths through a three-way control valve, the first path is led out of the containment vessel through the release valve and the isolation valve, and the second path is led into the containment vessel through the return pipe. Under the condition that the SGTR does not occur, adjusting a three-way control valve to control the first way to be normally open and the second way to be normally closed, and leading steam and water to the outside of the containment vessel for discharging according to the original design; under the condition that the SGTR accident happens, high-pressure radioactive steam and water are contained in the main steam pipeline, the isolation valve and the release valve are closed at the moment, the main steam pipeline is caused to jack the safety valve to enable the radioactive steam and water to flow out of the safety valve, the three-way control valve is adjusted to close the first way, the second way is opened, and the steam and water containing a large amount of radioactivity are discharged into pressure relief equipment in a containment through a return pipe.
Further, as a second preferable scheme, the release valve and the isolation valve are arranged outside the containment vessel, the safety valve is arranged in the containment vessel, the pressure relief device is arranged in the containment vessel and connected with one end of the return pipe, the other end of the return pipe is communicated with one end of the safety valve through the three-way control valve, the other end of the safety valve is communicated with the main steam pipeline, and the release valve and the isolation valve are sequentially fixed on the main steam pipeline and are both located at the downstream position of the safety valve in the fluid output direction. Under the condition that the SGTR does not occur, the first path is normally open, the second path is normally closed, and steam and water are led out of the containment and then discharged to external treatment equipment along the main steam pipeline; under the condition that an SGTR accident occurs, the isolation valve and the release valve are closed, the safety valve is jacked open due to the high pressure of the main steam pipeline, at the moment, the first path is controlled to be closed, the second path is controlled to be opened, and steam and water containing a large amount of radioactivity are discharged into pressure relief equipment in a containment.
Further, as a third preferable scheme, the safety valve, the release valve and the isolation valve are all arranged outside the containment vessel, the pressure relief device is arranged in the containment vessel and connected with one end of the return pipe, the other end of the return pipe is simultaneously connected with one end of the safety valve and one end of the release valve through the three-way control valve, the other end of the safety valve and the other end of the release valve are both communicated with the main steam pipeline, the isolation valve is fixed on the main steam pipeline, and the safety valve, the release valve and the isolation valve are sequentially distributed on the main steam pipeline in the fluid output direction. Different from the preferred scheme, the steam and water released by the release valve are led to the steam and water outlet of the safety valve. After the SGTR occurs, the release valve does not need to be closed, and radioactive steam and water released from the release valve are introduced into the pressure relief device in the containment.
Further, as a fourth preferable scheme, the isolation valve is arranged outside the containment vessel, the release valve is arranged inside or outside the containment vessel, the safety valve is arranged inside the containment vessel, the pressure relief device is arranged inside the containment vessel and connected with one end of the return pipe, the other end of the return pipe is simultaneously connected with one end of the safety valve and one end of the release valve through the three-way control valve, the other end of the safety valve and the other end of the release valve are both communicated with the main steam pipeline, the isolation valve is fixed on the main steam pipeline, and the safety valve, the release valve and the isolation valve are sequentially distributed on the main steam pipeline in the fluid output direction. The third difference between the scheme and the preferred scheme is that the safety valve can be built in the containment vessel or directly arranged on the steam generator, and the safety valve and the release valve can also be built in the containment vessel according to engineering requirements.
Further, the pressure relief device is a pressure stabilizer pressure relief box or a water pool, and the pressure relief device is arranged inside the containment.
The liquid part of the pressure relief equipment can lead water out of the containment vessel through a pipeline for cooling and drainage, so that the airborne radioactivity can be guaranteed to be retained in the containment vessel, and the volume and pressure of the water in the pressure relief equipment can be guaranteed not to exceed the limit.
In addition, for the four preferable schemes, any one of the following pressure relief devices can be matched as follows:
the first alternative is as follows: using the original pressure stabilizer to unload the pressure tank;
alternative two: a pool is arranged in an original containment;
alternative three: pressure relief and suppression equipment can be arranged independently;
the alternative is four: if the reactor containment is a closed double-layer containment, the radioactive emission can be controlled by a processing facility such as a radioactive filtering system of the double-layer containment and the like through introducing the radioactive emission into an annular cavity in the middle of the double-layer containment.
Further, a closed double-layer containment is adopted in the containment, and an annular cavity in the middle of the double-layer containment is used as a pressure relief device.
In the case that the containment is a closed double-layer containment, the radioactive emission can be controlled by a processing facility such as a radioactive filtering system of the double-layer containment annular cavity through introducing the radioactive emission into an annular cavity in the middle of the double-layer containment.
Furthermore, a pressure stabilizer pressure relief box or a water pool is arranged in the ring cavity and used as pressure relief equipment.
Compared with the prior art, the invention has the following advantages and beneficial effects:
the slow release device for the steam generator heat transfer pipe rupture accident utilizes the arrangement relationship among the steam generator, the main steam pipeline, the containment, the safety valve, the release valve and the isolation valve to carry out reasonable and feasible rearrangement, and introduces a large amount of radioactivity released by the release valve and the safety valve into the containment to control after the SGTR occurs, thereby perfecting the integrity of the barrier of the containment after the SGTR occurs, reducing the radioactive leakage risk caused by the containment bypass, reducing the negative influence of the radioactivity on the environment and people, and providing radiation safety guarantee for the related construction of a reactor.
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 schematic structural view of example 2;
FIG. 2 is a schematic structural view of embodiment 3;
FIG. 3 is a schematic structural view of example 4;
FIG. 4 is a schematic structural view of example 5.
Reference numbers and corresponding part names in the drawings: 1-containment vessel, 2-steam generator, 3-pressure relief equipment, 4-cavity, 5-return pipe, 6-safety valve, 7-three-way control valve, 8-release valve, 9-isolation valve and 10-main steam pipeline.
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 embodiment comprises a containment vessel 1 and a steam generator 2 arranged in the containment vessel 1, wherein a steam outlet of the steam generator 2 is connected with a main steam pipeline 10, the end part of the main steam pipeline 10 penetrates through the containment vessel 1 and then extends outwards, a slow release mechanism is arranged on the main steam pipeline 10, and the slow release mechanism comprises a pressure relief device 3, a return pipe 5, a safety valve 6, a three-way control valve 7, a release valve 8 and an isolation valve 9. On the basis that the reactor comprises a conventional two-loop system, the radioactive release of the SGTR accident is relieved by reasonably arranging a safety valve, a release valve and an isolation valve on a main steam pipeline and additionally arranging a three-way control valve, a return pipe and pressure relief equipment.
Aiming at the problem that when an SGTR accident occurs, a large amount of two-loop coolant containing radioactive substances passes through the last safety barrier of a reactor plant in a gaseous state or even in a liquid state and is directly discharged into the environment, if the reactor core is molten, a large amount of the radioactive substances generated by the melting can also be discharged into the environment through the crevasses, the applicant provides a slow release device for releasing a large amount of the radioactive substances due to the fracture of a heat transfer pipe of a steam generator 2, namely, the slow release device is used for introducing a large amount of radioactivity generated by the SGTR accident into a containment 1 to control, the integrity of the barrier of the containment 1 after the SGTR occurs is perfected, the radioactive leakage risk caused by the bypass of the containment 1 is reduced, the negative influence of the radioactivity on the environment and people is reduced, and the radiation safety guarantee is provided for the related construction of the reactor. The applicant controls the radioactive substances leaked from the steam generator 2 to flow back to the recovery processing equipment through different valve systems by arranging two pipelines communicated with the main steam pipeline 10, so that the radioactive consequences under the conditions of a benchmark accident and a serious accident are reduced, and the safety coefficient of the reactor is improved.
Example 2
As shown in fig. 1, in this embodiment, the slow release mechanism includes a pressure relief device 3, a return pipe 5, a safety valve 6, a three-way control valve 7, a release valve 8, and an isolation valve 9, the safety valve 6, the release valve 8, and the isolation valve 9 are all disposed outside the containment vessel 1, the pressure relief device 3 is disposed in the containment vessel 1 and is communicated with an outlet at one end of the three-way control valve 7 through the return pipe 5, an outlet at the other end of the three-way control valve 7 is communicated with an outlet of the original safety valve 6, an inlet of the three-way control valve 7 is connected with an outlet of the safety valve 6, an inlet of the safety valve 6 is communicated with a main steam pipeline. Further, after the pressure of water and steam in the steam generator 2 and the main steam pipeline 10 exceeds the protection pressure of the safety valve 6, the steam and the water are released from the safety valve 6, then enter the three-way control valve 7 and are divided into two paths, the first path is led out of the containment vessel 1, the second path is returned to the interior of the containment vessel 1 through the return pipe 5, and the pressure relief equipment 3 is connected to the tail end of the return pipe 5, and the pressure relief equipment 3 can relieve the pressure of high-temperature and high-pressure steam returned to the interior of the containment vessel 1 so as to facilitate the post-treatment of the radioactive filtering equipment; specifically, under the condition that the SGTR does not occur, the three-way control valve 7 enables the first path to be normally open and the second path to be normally closed, and the steam and water are led out of the containment vessel 1 and then discharged to external treatment equipment along the main steam pipeline 10; under the condition that an SGTR accident occurs, the isolation valve 9 and the release valve 8 are closed, the safety valve 6 is jacked open due to the high pressure of the main steam pipeline 10, at the moment, the three-way control valve 7 closes the first way, opens the second way, and discharges steam and water containing a large amount of radioactivity into pressure relief equipment in the containment vessel 1, so that the release of the large amount of radioactivity is practically eliminated.
Example 3
As shown in fig. 2, the present embodiment is different from embodiment 2 in that a safety valve 6 is built in the containment 1 or directly provided on the steam generator 2. This embodiment also virtually eliminates a significant amount of radioactivity release.
Example 4
As shown in fig. 3, the difference between this embodiment and embodiment 2 is that the steam and water released from the release valve 8 are led to the outlet of the safety valve 6 and are connected to the inlet of the three-way control valve 7. After the SGTR occurs, the release valve 8 does not need to be closed, and radioactive vapor and water released from the release valve 8 are introduced into the pressure relief device 3 in the containment vessel 1. This embodiment also virtually eliminates a significant amount of radioactivity release.
Example 5
As shown in fig. 4, the present embodiment is different from embodiment 3 in that the safety valve 6 is built into the containment 1, or is directly disposed on the steam generator 2, and the safety valve 6 and the release valve 7 can also be built into the containment 1 according to the engineering requirement. This embodiment also virtually eliminates a significant amount of radioactivity release.
Example 6
As shown in fig. 1 to 4, in this embodiment, on the basis of embodiments 1 to 5, the pressure relief device 3 is a pressure stabilizer pressure relief tank, a water pool or a dedicated pressure relief device, and the pressure relief device 3 is disposed inside the containment vessel 1. The pressure relief device 3 can convert high-temperature and high-pressure steam which enters the containment 1 again from the return pipe 5 into liquid at normal temperature and normal pressure, and the pressure relief device 3 is arranged in the containment 1 aiming at the single-layer containment 1, so that the radioactive filtering device can rapidly process the steam leaked due to the rupture of the heat transfer pipe in a relatively stable environment.
Example 7
As shown in fig. 1 to 4, in this embodiment, in addition to embodiments 1 to 5, the containment vessel 1 is a double-layer containment vessel, an annular cavity 4 between the double-layer containment vessel is used as a pressure relief device 3, or the pressure relief device 3 is arranged in the cavity 4. The pressure stabilizer pressure relief box or the water pool can convert high-temperature and high-pressure water vapor which enters the containment vessel 1 again from the return pipe 5 into liquid at normal temperature and normal pressure, and the pressure stabilizer pressure relief box or the water pool is arranged in the cavity 4 of the containment vessel 1 aiming at the double-layer containment vessel 1, so that the water vapor leaked due to the rupture of the heat transfer pipe can be rapidly processed conveniently by the radioactive filtering system in the cavity 4 in a relatively stable environment; in the cavity 4, the treated liquid part can be led out of the containment vessel 1 through a pipeline to be cooled and drained, so that airborne radioactivity can be retained in the containment vessel 1, and the volume of water in the pressure relief equipment and the pressure of the containment vessel 1 can be prevented from exceeding the limit.
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 (9)

1. A slow-release device that is used for steam generator heat-transfer pipe rupture accident, its characterized in that: the safety shell comprises a safety shell (1) of a steam generator and a slow release mechanism, wherein radioactive substances released from the safety shell (1) after a reactor SGTR accident occurs are introduced into the safety shell (1) by the slow release mechanism.
2. A slow-release device that is used for steam generator heat-transfer pipe rupture accident, its characterized in that: the safety shell is characterized in that a steam generator (2) is arranged inside the safety shell (1), a main steam pipeline (10) is connected to a steam outlet of the steam generator (2), the end portion of the main steam pipeline (10) penetrates through the safety shell (1) and extends outwards, the slow release mechanism is arranged on the main steam pipeline (10), and the slow release mechanism comprises a pressure release device (3), a return pipe (5), a safety valve (6), a three-way control valve (7), a release valve (8) and an isolation valve (9).
3. The slow release device for a steam generator heat transfer tube rupture event according to claim 1, wherein: the safety valve (6), the release valve (8) and the isolation valve (9) are all arranged outside the containment vessel (1), the pressure relief device (3) is arranged in the containment vessel (1) and is connected with one end of the return pipe (5), the other end of the return pipe (5) is communicated with one end of the safety valve (6) through the three-way control valve (7), and the other end of the safety valve (6) is communicated with the main steam pipeline (10); along the fluid output direction, a release valve (8) and an isolation valve (9) are sequentially fixed on the main steam pipeline (10) and are positioned at the downstream position of the safety valve (6).
4. The slow release device for a steam generator heat transfer tube rupture event according to claim 1, wherein: the safety device is characterized in that the release valve (8) and the isolation valve (9) are arranged outside the containment vessel (1), the safety valve (6) is arranged in the containment vessel (1), the pressure relief equipment (3) is arranged in the containment vessel (1) and is connected with one end of the return pipe (5), the other end of the return pipe (5) is communicated with one end of the safety valve (6) through the three-way control valve (7), the other end of the safety valve (6) is communicated with the main steam pipeline (10), and the release valve (8) and the isolation valve (9) are sequentially fixed on the main steam pipeline (10) and are located at the downstream position of the safety valve (6) in the fluid output direction.
5. The slow release device for a steam generator heat transfer tube rupture event according to claim 1, wherein: the safety valve (6), the release valve (8) and the isolation valve (9) are arranged outside the containment vessel (1), the pressure relief device (3) is arranged in the containment vessel (1) and is connected with one end of the return pipe (5), the other end of the return pipe (5) is connected with one end of the safety valve (6) and one end of the release valve (8) through the three-way control valve (7), the other end of the safety valve (6) and the other end of the release valve (8) are communicated with the main steam pipeline (10), the isolation valve (9) is fixed on the main steam pipeline (10), and in the fluid output direction, the safety valve (6), the release valve (8) and the isolation valve (9) are sequentially distributed on the main steam pipeline (10).
6. The slow release device for a steam generator heat transfer tube rupture event according to claim 1, wherein: the safety device is characterized in that the isolation valve (9) is arranged outside the containment vessel (1), the release valve (8) is arranged in or outside the containment vessel (1), the safety valve (6) is arranged in the containment vessel (1), the pressure relief device (3) is arranged in the containment vessel (1) and connected with one end of the return pipe (5), the other end of the return pipe (5) is simultaneously connected with one end of the safety valve (6) and one end of the release valve (8) through the three-way control valve (7), the other end of the safety valve (6) and the other end of the release valve (8) are communicated with the main steam pipeline (10), the isolation valve (9) is fixed on the main steam pipeline (10), and the safety valve (6), the release valve (8) and the isolation valve (9) are sequentially distributed on the main steam pipeline (10) in the fluid output direction.
7. The slow release device for the steam generator heat transfer tube rupture accident according to any one of claims 2 to 6, wherein: the pressure relief device (3) is a pressure stabilizer pressure relief box or a water pool, and the pressure relief device (3) is arranged inside the containment (1).
8. The slow release device for the steam generator heat transfer tube rupture accident according to any one of claims 2 to 6, wherein: and a closed double-layer containment is adopted in the containment (1), and an annular cavity in the middle of the double-layer containment is used as a pressure relief device (3).
9. The slow release device for a steam generator heat transfer tube rupture event according to claim 8, wherein: and a pressure stabilizer pressure relief box or a water pool is arranged in the ring cavity and used as pressure relief equipment (3).
CN201911150155.2A 2019-11-21 2019-11-21 Sustained release device for steam generator heat transfer tube rupture accident Pending CN110911024A (en)

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US20210313082A1 (en) * 2020-04-01 2021-10-07 Korea Atomic Energy Research Institute Steam generator accident mitigation system
US11823804B2 (en) * 2020-04-01 2023-11-21 Korea Atomic Energy Research Institute Steam generator accident mitigation system
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

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Application publication date: 20200324