CN113488214A - Natural circulation cooling method for steam on upper end enclosure of pressure vessel of nuclear power plant - Google Patents
Natural circulation cooling method for steam on upper end enclosure of pressure vessel of nuclear power plant Download PDFInfo
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- CN113488214A CN113488214A CN202110830446.7A CN202110830446A CN113488214A CN 113488214 A CN113488214 A CN 113488214A CN 202110830446 A CN202110830446 A CN 202110830446A CN 113488214 A CN113488214 A CN 113488214A
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- 238000001816 cooling Methods 0.000 title claims abstract description 77
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims abstract description 56
- 239000003381 stabilizer Substances 0.000 claims description 27
- 238000000034 method Methods 0.000 claims description 10
- 238000007599 discharging Methods 0.000 claims description 9
- 230000007246 mechanism Effects 0.000 claims description 8
- 238000005507 spraying Methods 0.000 claims description 6
- 238000004781 supercooling Methods 0.000 claims description 4
- 238000001704 evaporation Methods 0.000 abstract description 5
- 230000008020 evaporation Effects 0.000 abstract description 5
- 239000002918 waste heat Substances 0.000 abstract description 4
- 239000007789 gas Substances 0.000 description 6
- 230000008569 process Effects 0.000 description 4
- 239000007921 spray Substances 0.000 description 4
- 230000007704 transition Effects 0.000 description 4
- 238000002347 injection Methods 0.000 description 3
- 239000007924 injection Substances 0.000 description 3
- 239000007788 liquid Substances 0.000 description 3
- 239000002904 solvent Substances 0.000 description 3
- 239000000126 substance Substances 0.000 description 3
- 239000002826 coolant Substances 0.000 description 2
- 238000013461 design Methods 0.000 description 2
- 238000005516 engineering process Methods 0.000 description 2
- 239000000243 solution Substances 0.000 description 2
- 238000012546 transfer Methods 0.000 description 2
- 230000009286 beneficial effect Effects 0.000 description 1
- 230000033228 biological regulation Effects 0.000 description 1
- 238000005271 boronizing Methods 0.000 description 1
- 238000010276 construction Methods 0.000 description 1
- 230000010485 coping Effects 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 239000012530 fluid Substances 0.000 description 1
- 230000017525 heat dissipation Effects 0.000 description 1
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 238000011946 reduction process Methods 0.000 description 1
- 238000007789 sealing Methods 0.000 description 1
- 238000006467 substitution reaction Methods 0.000 description 1
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Classifications
-
- 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
-
- 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
- G21C15/02—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
In order to solve the problem that the safety of a reactor is endangered because the reactor cannot discharge waste heat due to the fact that a flash evaporation steam generation phenomenon occurs to an upper end socket in the prior art, the embodiment of the invention provides a natural circulation cooling method for the upper end socket of a pressure vessel of a nuclear power plant when steam exists, which comprises the following steps: controlling the main system to cool and reduce the pressure; checking whether the upper end enclosure of the pressure container has a water level or not, and controlling the main system to boost pressure if the upper end enclosure of the pressure container has no water level; if the upper end enclosure of the pressure container has a water level, continuing cooling; checking whether the temperature of the hot section is less than 177 ℃ and the pressure of the main system is less than 2.7 MPa; if yes, cooling the main system to a cold shutdown stack; if not, controlling the main system to continuously cool and reduce the pressure; checking whether the upper end enclosure of the pressure container has a water level, if the upper end enclosure of the pressure container has no water level, controlling the pressure container to be boosted until the upper end enclosure has the water level, and cooling the main system until the main system is cooled to be in cold shutdown; cooling the main system dead zone; checking whether the temperature of the system is less than 90 ℃, if so, completely relieving the pressure of the main system; if not, cooling the main system to the cold shutdown.
Description
Technical Field
The invention relates to a natural circulation cooling method for a nuclear power plant pressure vessel upper end socket in the presence of steam.
Background
Hualongyi (HPR1000) is an advanced million kilowatt-level pressurized water reactor nuclear power model with complete and independent intellectual property rights in China, absorbs nuclear power design, construction and operation experience of China for thirty years and international advanced nuclear power design characteristics, adopts active and passive safety technology, and achieves the advanced level of the international third-generation nuclear power technology in technical indexes and safety indexes.
After the Hualongyi nuclear power plant has a non-loss-of-coolant accident and triggers the reactor to be stopped emergently, an operator controls and relieves the accident consequence according to the emergency accident regulations, and if the accident cannot be repaired in the hot shutdown state of the nuclear power plant, the reactor needs to be cooled to the cold shutdown state, so that the fault treatment is carried out. During the accident handling process, the reactor loop system will be in a natural circulation operating state if the reactor coolant pump is not available.
The reactor pressure vessel upper head area has a flow dead zone, a reactor loop system cools and reduces pressure in the accident process, and the cooling rate of the upper head dead zone is lower than that of the loop system. Because the cooling capacity of the dead zone of the upper end enclosure is insufficient, the temperature of the fluid of the upper end enclosure reaches the saturation temperature in the depressurization process of a loop system of the reactor, and the phenomenon of flash evaporation and steam generation occurs. If the steam yield is large, the liquid level in the pressure container is lower than the upper surface of the heat pipe section, steam enters the heat pipe section of the loop system and then flows into the heat transfer pipe of the steam generator, and the steam is gathered at the top of the heat transfer pipe of the steam generator, so that the natural circulation flow is stopped, the waste heat of the reactor core of the reactor cannot be discharged, and the safety of the reactor is endangered.
Disclosure of Invention
In order to solve the technical problem that the safety of a reactor is endangered due to the fact that waste heat cannot be discharged by the reactor because the phenomenon of flash evaporation and steam generation of an upper end enclosure exists in the prior art, the embodiment of the invention provides a natural circulation cooling method for the upper end enclosure of a pressure vessel of a nuclear power plant when steam exists.
The embodiment of the invention is realized by the following technical scheme:
a natural circulation cooling method for a nuclear power plant pressure vessel upper end socket in the presence of steam comprises the following steps:
controlling the main system to cool and reduce the pressure;
checking whether the upper end enclosure of the pressure container has a water level or not, and controlling the main system to boost pressure if the upper end enclosure of the pressure container has no water level; if the upper end enclosure of the pressure container has water level, then
Checking whether the temperature of the hot section is less than 177 ℃ and the pressure of the main system is less than 2.7 MPa; if yes, cooling the main system to a cold shutdown stack; if not, controlling the main system to cool and reduce the pressure;
checking whether the upper end enclosure of the pressure container has a water level, if the upper end enclosure of the pressure container has no water level, controlling the pressure container to be boosted until the upper end enclosure has the water level, and cooling the main system until the main system is cooled to be in cold shutdown;
cooling the main system dead zone;
checking whether the temperature of the system is less than 90 ℃, if so, completely relieving the pressure of the main system; if not, cooling the main system to the cold shutdown.
Further, the method further comprises: and before the temperature and the pressure are reduced, the water level of the voltage stabilizer is controlled so that the voltage stabilizer is used for containing bubbles generated by the main system.
Further, controlling the water level of the pressurizer to enable the pressurizer to contain bubbles generated by the main system comprises: the water level of the voltage stabilizer is controlled to be 12% -22% of the measuring range.
Further, the main system is controlled to cool and reduce the pressure, and the method comprises the following steps: maintaining the cooling rate of the cold section of the main system to be less than 56 ℃/h and maintaining the supercooling degree of the reactor core outlet in the pressure vessel to be more than 20 ℃.
Further, control main system and carry out cooling step-down, still include: and (4) reducing the pressure by using an auxiliary spraying or pressure stabilizer safety valve.
Further, control main system and carry out cooling step-down, still include: the water level of the voltage stabilizer is controlled to be 12% -86% of the measuring range.
Further, cooling the primary system to the cold shutdown stack includes: and cooling the reactor main system by using the surplus exhaust system.
Further, cooling the main system dead zone, comprising: and cooling the upper end enclosure of the pressure vessel by using a cooling fan of the control rod driving mechanism, and discharging and cooling the U-shaped tube area of the steam generator through steam.
Compared with the prior art, the invention has the following advantages and beneficial effects:
according to the natural circulation cooling method for the upper end enclosure of the pressure vessel of the nuclear power plant in the embodiment of the invention, the water level of the upper end enclosure is controlled to prevent water vapor in the upper end enclosure from entering a hot section in the process of cooling and depressurizing a main system; the temperature and pressure reduction process of the main system is realized through a corresponding processing mode; therefore, hot gas in the upper sealing head is prevented from entering the steam generator to block the circulation cooling of the steam generator.
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 diagram of a pressure vessel cooling system.
FIG. 2 is a schematic flow chart of a natural circulation cooling method when steam exists at an upper end enclosure of a pressure vessel of a nuclear power plant.
Reference numbers and corresponding part names in the drawings:
1-a pressure vessel; 2-control rod drive mechanism; 3-a voltage stabilizer; 4-a steam generator; 5-main pump; 6-a pressurizer safety valve; 7-pressure relief box; 8, supplying water by an auxiliary spraying system; 9-a spray valve; 10-a waste heat discharge pump; 11-residual heat removal heat exchanger; 12-a water supply channel; 13-a steam channel; 14-chemical and solvent control systems; 15-a cold section; 16-a transition section; 17-safety injection box; 18-hot section; 19-upper end enclosure; 20-dead zone; 21-hot gas zone.
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.
The single-point temperature in the embodiment of the invention represents the single-point temperature +/-2 ℃; the single point pressure indicates the single point pressure of +/-0.2 MPa.
Examples
Referring to fig. 1, the pressure vessel cooling system includes a pressure vessel 1, a control rod drive mechanism 2, a pressure stabilizer 3, a steam generator 4, a main pump 5, a pressure stabilizer safety valve 6, a pressure relief tank 7, an auxiliary spray 8, a spray valve 9, a residual heat removal pump 10, a residual heat removal heat exchanger 11, a chemical and solvent control system 14, and a safety injection tank 17.
The main system comprises a pressure vessel 1, a control rod driving mechanism 2, a voltage stabilizer 3 and a steam generator 4; the pressure container 1 is respectively communicated with the voltage stabilizer 3 through a hot section 18; the pressure vessel is communicated with the steam generator 4 through the hot section; the steam generator is connected with the main pump 5 through a transition section 16 and then communicated with the cold section 15; the cold leg returns to the pressure vessel.
The hot section 18 is also connected with a safety injection tank 17.
The primary system pressure refers to the internal pressure of the system consisting of the pressure vessel 1, the control rod drive mechanism 2, the pressurizer 3, the steam generator 4, and the main pump 5, as well as the hot leg, the transition leg, and the cold leg.
The residual heat discharging system comprises a residual heat discharging pump 10 and a residual heat discharging heat exchanger 11, and the hot section 18 is communicated with the cold section 15 sequentially through the residual heat discharging pump and the residual heat discharging heat exchanger 11.
The pressure stabilizer 3 is communicated with the pressure relief tank 7 through a safety valve 6 of the pressure stabilizer.
Wherein, the hot section, the cold section and the transition section are pipelines.
The cold section 15 is also communicated with a hot air area 21 in the voltage stabilizer 3 through a spray valve 9; the hot gas zone 21 is also sprayed with water 8 by means of an auxiliary spraying system.
The upper part of the pressure vessel 1 has an upper closure head 19, i.e. a dead space 20 section. And the chemical and solvent control system is used for conveying the boronizing liquid to the pressure container and the interior of the pressure stabilizer 3 to cool the pressure container and the pressure stabilizer.
The steam generator 4 cools the main system through a U-shaped pipe of the steam generator, and the steam generator discharges steam through a steam channel 13; the steam generator injects water into the steam generator through the feed water channel 12.
When the whole system needs to be cooled, safety accidents are easily caused by flash evaporation caused by steam in the upper end enclosure.
Specifically, when steam exists in the upper head 19 to initiate flash evaporation, if the liquid level in the pressure vessel is lower than the hot section 18, hot gas in the upper head enters the steam generator 4 along the hot section and is accumulated in the upper part of the steam generator 4, so that the gas pressure in the upper part of the steam generator is increased; the air pressure at the upper part of the steam generator is increased, so that the circulation of the U-shaped pipe area in the steam generator is influenced, and even the air inlet and the heat dissipation of the pressure vessel 1 cannot be realized through the U-shaped pipe of the steam generator, thereby causing safety accidents.
In order to deal with the accident, the inventor gives the following coping method by taking the idea of preventing the hot gas in the upper head from entering the steam generator as a whole when the upper head has the steam.
Referring to fig. 2, a natural circulation cooling method for a nuclear power plant pressure vessel with steam on the upper head comprises the following steps:
controlling the main system to cool and reduce the pressure;
checking whether the upper end enclosure of the pressure container has a water level or not, and controlling the main system to boost pressure if the upper end enclosure of the pressure container has no water level; if the upper end enclosure of the pressure container has water level, then
Checking whether the temperature of the hot section is less than 177 ℃ and the pressure of the main system is less than 2.7 MPa; if yes, cooling the main system to a cold shutdown stack; if not, controlling the main system to cool and reduce the pressure;
checking whether the upper end enclosure of the pressure container has a water level, if the upper end enclosure of the pressure container has no water level, controlling the pressure container to be boosted until the upper end enclosure has the water level, and cooling the main system until the main system is cooled to be in cold shutdown;
cooling the main system dead zone;
checking whether the temperature of the system is less than 90 ℃, if so, completely relieving the pressure of the main system; if not, cooling the main system to the cold shutdown.
Further, the method further comprises: the water level of the pressurizer is controlled so that the pressurizer can contain bubbles generated by the main system.
Further, controlling the water level of the pressurizer to enable the pressurizer to contain bubbles generated by the main system comprises: the water level of the voltage stabilizer is controlled to be 12% -22% of the measuring range.
Further, the main system is controlled to cool and reduce the pressure, and the method comprises the following steps: maintaining the cooling rate of the cold section of the main system to be less than 56 ℃/h and maintaining the supercooling degree of the reactor core outlet in the pressure vessel to be more than 20 ℃.
Further, control main system and carry out cooling step-down, still include: and (4) reducing the pressure by using an auxiliary spraying or pressure stabilizer safety valve.
Further, control main system and carry out cooling step-down, still include: the water level of the voltage stabilizer is controlled to be 12% -86% of the measuring range.
Further, cooling the primary system to the cold shutdown stack includes: and cooling the reactor main system by using the surplus exhaust system.
Further, cooling the main system dead zone, comprising: and cooling the upper end enclosure of the pressure vessel by using a cooling fan of the control rod driving mechanism, and discharging and cooling the U-shaped tube area of the steam generator through steam.
In the specific operation, the following steps can be adopted:
step 1: a pressurizer water level is established that can accommodate bubble generation. Checking that the water level of the voltage stabilizer ranges from 12% to 22%, and adjusting the upper charging and lower discharging flow according to the requirement; the water level of the voltage stabilizer is controlled to be switched to manual.
Step 2: the main system is cooled and depressurized. Maintaining the cooling rate of the cold section of the main system to be less than 56 ℃/h and maintaining the supercooling degree of the outlet of the reactor core to be more than 20 ℃; and (4) reducing the pressure by using an auxiliary spraying or pressure stabilizer safety valve.
And step 3: and controlling the water level of the voltage stabilizer to be between 12% and 86% of the range.
And 4, step 4: and checking a water level indicating system of the pressure container to indicate that the upper seal head has a water level. And if the upper end enclosure of the pressure container has no water level, boosting the main system to enable the pressure container water level indicating system to indicate that the upper end enclosure has the water level, and returning to the step 2.
And 5: and (3) if the temperature of the hot section is less than 177 ℃ and the pressure of the main system is less than 2.7MPa, putting the reactor into the residual discharge system, cooling the main system of the reactor by using the residual discharge system, and otherwise, returning to the step (2).
And 6, cooling the main system to the cold shutdown.
And 7: cooling the main system dead zone. And cooling the upper end enclosure of the pressure vessel by using a cooling fan of the control rod driving mechanism, and cooling all areas of the U-shaped tubes of the steam generator through steam discharge. And (6) checking the water level indicating system of the pressure container to indicate that the upper end enclosure has the water level, and if the water level does not exist, controlling the main system to boost the pressure and returning to the step 6.
And 8: it is determined whether the main system requires full pressure relief. The temperature of the whole main system is less than 90 ℃, and the main system needs to be completely decompressed; and if the temperature of the whole main system is greater than or equal to 90 ℃, returning to the step 6.
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 (8)
1. A natural circulation cooling method for a nuclear power plant pressure vessel upper end socket in the presence of steam is characterized by comprising the following steps:
controlling the main system to cool and reduce the pressure;
checking whether the upper end enclosure of the pressure container has a water level, and if the upper end enclosure of the pressure container has no water level, controlling the main system to boost the pressure so that the upper end enclosure of the pressure container has the water level; if the upper end enclosure of the pressure container has water level, then
Checking whether the temperature of the hot section is less than 177 ℃ and the pressure of the main system is less than 2.7 MPa; if yes, cooling the main system to a cold shutdown stack; if not, controlling the main system to cool and reduce the pressure;
checking whether the upper end enclosure of the pressure container has a water level, if the upper end enclosure of the pressure container has no water level, controlling the pressure container to be boosted until the upper end enclosure has the water level, and cooling the main system until the main system is cooled to be in cold shutdown;
cooling the main system dead zone;
checking whether the temperature of the system is less than 90 ℃, if so, completely relieving the pressure of the main system; if not, cooling the main system to the cold shutdown.
2. The natural circulation cooling method for the upper head of the nuclear power plant pressure vessel with steam as claimed in claim 1, wherein the step of controlling the main system to reduce the temperature and pressure further comprises: and before the temperature and the pressure are reduced, the water level of the voltage stabilizer is controlled so that the voltage stabilizer is used for containing bubbles generated by the main system.
3. The method of claim 2, wherein controlling the water level of the pressurizer to contain bubbles generated by the primary system comprises: the water level of the voltage stabilizer is controlled to be 12% -22% of the measuring range.
4. The natural circulation cooling method for the upper head of the nuclear power plant pressure vessel with steam as claimed in claim 1, wherein the step of controlling the main system to reduce the temperature and pressure comprises the following steps: maintaining the cooling rate of the cold section of the main system to be less than 56 ℃/h and maintaining the supercooling degree of the reactor core outlet in the pressure vessel to be more than 20 ℃.
5. The natural circulation cooling method for the upper head of the nuclear power plant pressure vessel with steam as claimed in claim 1, wherein the natural circulation cooling method comprises the following steps: control main system and carry out cooling step-down, still include: and (4) reducing the pressure by using an auxiliary spraying or pressure stabilizer safety valve.
6. The natural circulation cooling method for the upper head of the nuclear power plant pressure vessel with steam as claimed in claim 1, wherein the natural circulation cooling method comprises the following steps: control main system and carry out cooling step-down, still include: the water level of the voltage stabilizer is controlled to be 12% -86% of the measuring range.
7. The natural circulation cooling method for the upper head of the nuclear power plant pressure vessel with steam as claimed in claim 1, wherein the natural circulation cooling method comprises the following steps: cooling a primary system to a cold shutdown stack, comprising: and cooling the reactor main system by using the surplus exhaust system.
8. The natural circulation cooling method for the upper head of the nuclear power plant pressure vessel with steam as claimed in claim 1, wherein the natural circulation cooling method comprises the following steps: cooling the main system dead zone, comprising: and cooling the upper end enclosure of the pressure vessel by using a cooling fan of the control rod driving mechanism, and discharging and cooling the U-shaped tube area of the steam generator through steam.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202110830446.7A CN113488214B (en) | 2021-07-22 | 2021-07-22 | Natural circulation cooling method for upper end socket of pressure vessel of nuclear power plant when steam exists |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202110830446.7A CN113488214B (en) | 2021-07-22 | 2021-07-22 | Natural circulation cooling method for upper end socket of pressure vessel of nuclear power plant when steam exists |
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| Publication Number | Publication Date |
|---|---|
| CN113488214A true CN113488214A (en) | 2021-10-08 |
| CN113488214B CN113488214B (en) | 2024-01-23 |
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Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB190824674A (en) * | 1908-11-17 | 1909-04-15 | Charles Reeves Mills | Improvements in Safety Valves. |
| WO2016011569A1 (en) * | 2014-07-24 | 2016-01-28 | 哈尔滨工程大学 | Containment cooling system, and containment and reactor pressure vessel joint cooling system |
| 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 |
| CN111755139A (en) * | 2020-04-20 | 2020-10-09 | 中国核电工程有限公司 | Design method of pressure vessel stack top exhaust control strategy under accident condition |
| CN111753394A (en) * | 2020-05-20 | 2020-10-09 | 中国核电工程有限公司 | Design method for rapid cooling function debugging of primary circuit of advanced pressurized water reactor nuclear power plant |
-
2021
- 2021-07-22 CN CN202110830446.7A patent/CN113488214B/en active Active
Patent Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB190824674A (en) * | 1908-11-17 | 1909-04-15 | Charles Reeves Mills | Improvements in Safety Valves. |
| WO2016011569A1 (en) * | 2014-07-24 | 2016-01-28 | 哈尔滨工程大学 | Containment cooling system, and containment and reactor pressure vessel joint cooling system |
| CN111755139A (en) * | 2020-04-20 | 2020-10-09 | 中国核电工程有限公司 | Design method of pressure vessel stack top exhaust control strategy under accident condition |
| 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 |
| CN111753394A (en) * | 2020-05-20 | 2020-10-09 | 中国核电工程有限公司 | Design method for rapid cooling function debugging of primary circuit of advanced pressurized water reactor nuclear power plant |
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| CN113488214B (en) | 2024-01-23 |
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