CN117095840B - Self-supporting passive waste heat discharging system of floating nuclear power station - Google Patents
Self-supporting passive waste heat discharging system of floating nuclear power station Download PDFInfo
- Publication number
- CN117095840B CN117095840B CN202310714834.8A CN202310714834A CN117095840B CN 117095840 B CN117095840 B CN 117095840B CN 202310714834 A CN202310714834 A CN 202310714834A CN 117095840 B CN117095840 B CN 117095840B
- Authority
- CN
- China
- Prior art keywords
- water tank
- nuclear power
- waste heat
- isolation valve
- cooling water
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Active
Links
- 239000002918 waste heat Substances 0.000 title claims abstract description 39
- 238000007599 discharging Methods 0.000 title claims abstract description 14
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Chemical compound O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims abstract description 89
- 239000000498 cooling water Substances 0.000 claims abstract description 40
- 238000001816 cooling Methods 0.000 claims abstract description 29
- 238000002955 isolation Methods 0.000 claims description 44
- 238000004891 communication Methods 0.000 claims description 8
- 229910000831 Steel Inorganic materials 0.000 claims description 2
- 239000000843 powder Substances 0.000 claims description 2
- 239000010959 steel Substances 0.000 claims description 2
- 230000009286 beneficial effect Effects 0.000 abstract description 4
- 230000002035 prolonged effect Effects 0.000 abstract description 4
- 239000002826 coolant Substances 0.000 description 4
- 230000000694 effects Effects 0.000 description 4
- 230000001174 ascending effect Effects 0.000 description 2
- 238000006243 chemical reaction Methods 0.000 description 2
- 238000013461 design Methods 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 230000005484 gravity Effects 0.000 description 2
- 238000000034 method Methods 0.000 description 2
- 238000012986 modification Methods 0.000 description 2
- 230000004048 modification Effects 0.000 description 2
- 230000008569 process Effects 0.000 description 2
- 238000011084 recovery Methods 0.000 description 2
- 239000003381 stabilizer Substances 0.000 description 2
- 230000001502 supplementing effect Effects 0.000 description 2
- 230000008859 change Effects 0.000 description 1
- 238000009833 condensation Methods 0.000 description 1
- 230000005494 condensation Effects 0.000 description 1
- 238000009795 derivation Methods 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 238000009434 installation Methods 0.000 description 1
- 239000007788 liquid Substances 0.000 description 1
- 230000007774 longterm Effects 0.000 description 1
- 230000008520 organization Effects 0.000 description 1
- 238000010992 reflux Methods 0.000 description 1
- 238000011160 research Methods 0.000 description 1
- 239000013589 supplement Substances 0.000 description 1
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
- G21C15/14—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices from headers; from joints in ducts
-
- 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
- G21C15/24—Promoting flow of the coolant
- G21C15/257—Promoting flow of the coolant using heat-pipes
Landscapes
- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
The invention discloses a self-supporting passive waste heat discharging system of a floating nuclear power station, which comprises the following components: the inlet of the C-shaped pipe heat exchanger is communicated with one end of a waste heat discharge pipeline, the other end of the waste heat discharge pipeline is communicated with a steam generator or a main steam pipeline, the C-shaped pipe heat exchanger is arranged in a cooling water tank, a plurality of open thermosiphons are fixedly arranged on the upper surface of the cooling water tank, and the bottom ends of the open thermosiphons penetrate through the upper surface of the cooling water tank and are communicated with the cooling water tank; the water vapor formed by water heat exchange in the cooling water tank is discharged from the open type thermosiphon, and partial water vapor is condensed and can flow back into the water tank during discharge, so that the beneficial effect of self-holding of the water quantity of the cooling water tank is realized, the consumption speed of the water quantity of the water tank during nuclear power station accidents is greatly reduced, the cooling capacity is improved, the sustainable cooling time is prolonged, and the passive waste heat discharge requirement under the limited application environment of the space of the floating nuclear power station is met.
Description
Technical Field
The invention belongs to the field of passive waste heat derivation of reactors, and particularly relates to a self-supporting passive waste heat discharging system of a floating nuclear power station.
Background
After the international atomic energy organization first proposes the passive design concept, various research institutions worldwide invest a large amount of resources to realize the effective application of passive safety design in the field of nuclear engineering. The passive waste heat discharging system (the waste heat discharging system of the reactor core when the system of the nuclear power plant fails) is typically designed to take a large cooling water tank as a hot well, and realize the waste heat of the reactor core in a passive driving mode such as gravity, phase change and the like, and typical applications include an AP1000 advanced reactor designed by western house corporation in America, a SMART modularized integrated pressurized water reactor proposed by Korean atomic energy institute, a Hualong first third generation nuclear power unit independently developed in China and the like. However, the passive waste heat removal system using a large cooling water tank as a heat sink has the following drawbacks:
1. The operation capacity of the waste heat discharging system is limited by the water content of the water tank, the water content in the water tank is continuously consumed after an accident occurs, the cooling capacity is gradually reduced, and long-term cooling of the reactor core is difficult to realize under the extreme accident working condition.
2. In order to ensure the cooling capacity of the waste heat discharging system, the nuclear power station needs to be provided with a large-capacity cooling water tank, and the land nuclear power station has sufficient water tank installation space, but the floating nuclear power station has limited space and cannot be provided with a large-capacity water tank.
In view of this, how to provide a passive waste heat removal system suitable for a floating nuclear power plant is a technical problem that needs to be solved by those skilled in the art.
Disclosure of Invention
The invention aims to provide a self-supporting passive waste heat discharging system of a floating nuclear power station, which solves the problems, realizes the self-supporting of water for a cooling water tank, reduces the volume of the cooling water tank and meets the passive waste heat discharging requirement of the floating nuclear power station under the limited space condition.
In order to achieve the above object, the present invention provides the following solutions: a self-sustaining passive waste heat removal system for a floating nuclear power plant, the floating nuclear power plant including a steam generator in communication with a main steam line and a main feedwater line, comprising: the inlet of the C-shaped pipe heat exchanger is communicated with one end of a waste heat discharge pipeline, and the other end of the waste heat discharge pipeline is communicated with the steam generator or the main steam pipeline; the outlet of the C-shaped pipe heat exchanger is communicated with one end of a condensed water pipeline, and the other end of the condensed water pipeline is communicated with the steam generator or the main water supply pipeline; the C-shaped pipe heat exchanger is arranged in the cooling water tank; an open type thermosiphon (the open type thermosiphon is a type of heat pipe, the structure is simple, a liquid suction core is not arranged in the open type thermosiphon, and the recovery of internal working media is realized based on gravity and capillary effect), a plurality of open type thermosiphons are provided, a plurality of open type thermosiphons are fixedly arranged on the upper surface of the cooling water tank, and the bottom end of the open type thermosiphon penetrates through the upper surface of the cooling water tank and is communicated with the cooling water tank; the first isolation valve is arranged on the waste heat discharge pipeline; and the second isolation valve is arranged on the condensed water pipeline.
The beneficial effects of the invention are as follows:
The waste heat of the nuclear power station reactor core can be discharged into the C-shaped tube heat exchanger from the waste heat discharge pipeline in the form of steam through the steam generator, and condensed water is formed to flow back into the steam generator after the water tank is cooled; the water vapor formed by water heat exchange in the cooling water tank is discharged from the open type thermosiphon, partial water vapor is condensed during discharge and can flow back into the water tank, the beneficial effect of self-supporting water quantity of the cooling water tank is realized, the consumption speed of the water quantity of the cooling water tank during nuclear power station accidents is greatly reduced, the cooling capacity is improved, the sustainable cooling time is prolonged, the water capacity requirement of the cooling water tank is reduced, and the passive waste heat discharge requirement of the floating nuclear power station under the space limited application environment is met.
Further, a third isolation valve is arranged on the main steam pipeline, and a fourth isolation valve is arranged on the main water supply pipeline.
When the nuclear power station normally operates, the first isolation valve and the second isolation valve are closed, and the third isolation valve and the fourth isolation valve are opened. The main steam pipeline is used for transmitting steam generated by the steam generator to the loop system for energy conversion under the normal working condition of the nuclear power station, and the main water supply pipeline is used for supplementing water for the steam generator under the normal working condition of the nuclear power station. When an accident occurs, the third and fourth isolation valves are closed, the first and second isolation valves are opened, and at the moment, all water vapor generated by the core waste heat in the steam generator enters the C-shaped pipe heat exchanger from the waste heat discharge pipeline, and condensed water is formed after heat exchange in the water tank and flows back to the steam generator to form steam-condensed water circulation.
Further, when the other end of the waste heat discharging pipeline is communicated with the main steam pipeline, the communication position is positioned between the third isolation valve and the steam generator, and when the other end of the condensed water pipeline is communicated with the main water feeding pipeline, the communication position is positioned between the fourth isolation valve and the steam generator.
Adopt above-mentioned technical scheme's aim at guarantees that when third isolation valve and fourth isolation valve closed, the steam in the steam generator can get into waste heat recovery pipeline smoothly and the condenser pipe can flow back to steam generator in going into smoothly.
Further, the steam generator further comprises a check valve, wherein the check valve is arranged on the condensed water pipeline and is positioned between the second isolation valve and the steam generator. The check valve functions to prevent the condensed water from flowing backward from a position near the steam generator to a position near the cooling water tank.
Further, the cooling tower comprises a cooling tower body, a plurality of open thermosiphons are arranged in the cooling tower body and close to the bottom of the cooling tower body, the bottom of the cooling tower body is provided with a cold air inlet, and the top of the cooling tower body is provided with a hot air outlet.
The cooling tower has the advantages that the condensation effect of the open type thermosiphon is further improved, water vapor generated by the cooling water tank quickly forms condensed water in the open type thermosiphon to flow back to the cooling water tank, the water tank water self-holding effect is further enhanced, the consumption speed of the water tank water in the nuclear power station accident is reduced, the cooling capacity is improved, the sustainable cooling time is prolonged, and the water capacity requirement of the cooling water tank is reduced.
Further, one or more of a silk screen, a groove and a steel powder sintered layer are arranged on the inner side wall of the open type thermosiphon.
The structure is arranged on the inner side wall of the open type thermosiphon, so that the heat exchange speed of the open type thermosiphon can be increased, the capturing capacity of steam can be improved, the reflux capacity of condensed water can be improved by utilizing the enhancement of capillary effect, and the self-sustaining characteristic of the cooling water tank can be further enhanced.
Drawings
For a clearer description of an embodiment of the invention or of the solutions of the prior art, the drawings that are needed in the embodiment will be briefly described, it being obvious that the drawings in the following description are only some embodiments of the invention, and that other drawings can be obtained according to these drawings without inventive effort for a person skilled in the art:
FIG. 1 is a system block diagram of the present invention;
FIG. 2 is a schematic diagram of the operation of the present open thermosiphon;
Wherein: a 1-C type pipe heat exchanger, a 2-cooling water tank, a 3-cooling tower, a 4-cold air inlet, a 5-hot air outlet, a 6-open thermosiphon, a 7-inlet, an 8-outlet, a 9-first isolation valve, a 10-check valve, an 11-waste heat discharge pipe, a 12-condensed water pipe, a 13-second isolation valve, a 14-fourth isolation valve, a 15-third isolation valve, a 16-reactor core, a 17-steam generator, an 18-coolant ascending channel, a 19-pressure stabilizer and a 20-main pump.
Detailed Description
The following description of the embodiments of the present invention will be made clearly and completely with reference to the accompanying drawings, in which it is apparent that the embodiments described are only some embodiments of the present invention, but not all embodiments. All other embodiments, which can be made by those skilled in the art based on the embodiments of the invention without making any inventive effort, are intended to be within the scope of the invention.
In order that the above-recited objects, features and advantages of the present invention will become more readily apparent, a more particular description of the invention will be rendered by reference to the appended drawings and appended detailed description.
The embodiment provides a passive waste heat removal system of floating nuclear power station that can be self-sustaining, and floating nuclear power station includes steam generator 17, and steam generator 17 and main steam pipeline and main feed water pipeline intercommunication include: the heat pump type heat pump device comprises a C-shaped pipe heat exchanger 1, wherein an inlet 7 of the C-shaped pipe heat exchanger 1 is communicated with one end of a waste heat discharge pipeline 11, and the other end of the waste heat discharge pipeline 11 is communicated with a main steam pipeline; the outlet 8 of the C-shaped tube heat exchanger 1 is communicated with one end of a condensate water pipeline 12, and the other end of the condensate water pipeline 12 is communicated with a main water supply pipeline; the cooling water tank 2, the C-shaped tube heat exchanger 1 is arranged in the cooling water tank 2; the plurality of open thermosiphons 6 are arranged on the upper surface of the cooling water tank 2, and the bottom ends of the open thermosiphons 6 penetrate through the upper surface of the cooling water tank 2 and are communicated with the cooling water tank 2; the first isolation valve 9 is arranged on the waste heat discharge pipeline 11; and a second isolation valve 13, the second isolation valve 13 being disposed on the condensed water pipe 12.
In this embodiment, a third isolation valve 15 is provided on the main steam line and a fourth isolation valve 14 is provided on the main feedwater line.
In the present embodiment, the connection between the other end of the residual heat removal pipe 11 and the main steam pipe is located between the third isolation valve 15 and the steam generator 17, and the connection between the other end of the condensed water pipe 12 and the main water supply pipe is located between the second isolation valve 14 and the steam generator 17.
In the present embodiment, a check valve 10 is further included, and the check valve 10 is disposed on the condensed water pipe 12 between the second isolation valve 13 and the steam generator 17. The check valve 10 functions to prevent the condensed water from flowing backward from a position near the steam generator 17 to a position near the cooling water tank 2.
In this embodiment, the cooling tower 3 is further comprised of a plurality of open thermosiphons 6 disposed within the cooling tower 3 and positioned near the bottom of the cooling tower 3, the cooling tower 3 having a cool air inlet 4 at the bottom and a hot air outlet 5 at the top.
In this embodiment, the inner side wall of the open thermosiphon 6 is provided with a screen and grooves.
The specific working process is as follows:
Under normal working conditions, the coolant working medium enters the bottom of the steam generator 17 through the coolant ascending channel 18 after absorbing nuclear fission heat energy in the reactor core 16, exchanges heat with water in the secondary side of the steam generator 17, is pumped into the reactor core 16 by the main pump 20, and is used for maintaining the operating pressure by the pressure stabilizer 19. The steam generator 17 generates water steam after secondary side heat exchange, the water steam is transmitted to a loop system along a main steam pipeline to perform energy conversion, and meanwhile, the main water supply pipeline continuously supplements water into the steam generator 17. In this condition, the first isolation valve 9 and the second isolation valve 13 are closed, and the third isolation valve 15 and the fourth isolation valve 14 are opened.
Under the accident condition, the first isolation valve 9 and the second isolation valve 13 are opened, and the third isolation valve 15 and the fourth isolation valve 14 are closed. The coolant working medium still exchanges heat with water in the secondary side of the steam generator 17 after absorbing the residual heat of the reactor core 16, water vapor generated in the secondary side of the steam generator 17 after the heat exchange enters the C-shaped tube heat exchanger 1 through the residual heat discharge pipeline 11 and exchanges heat in the cooling water tank 2, condensed water is formed after the heat exchange, and flows back to the steam generator 17 along the condensing pipe, and at the moment, the water evaporated by the steam generator 17 and the water refluxed by the condensed water form a circulation without supplementing water to the steam generator 17. The water in the cooling water tank 2 exchanges heat with the C-tube heat exchanger 1 to form water vapor, and rises into the open thermosiphon 6. Part of water vapor in the open type thermosiphon 6 is condensed to form condensed water and flows back into the cooling water tank 2, meanwhile, cold air enters the cooling tower 3 from the cold air inlet 4 of the cooling tower 3 to cool the open type thermosiphon 6, and the condensed water becomes hot air after heat exchange and is discharged from the hot air outlet 5. During the process, part of the water vapor cannot be effectively cooled in the open thermosiphon 6, and is discharged from the hot air outlet 5 together with the hot air.
The invention provides a self-supporting passive residual heat discharging system of a floating nuclear power station, which can discharge the residual heat of the nuclear power station core into a C-shaped pipe heat exchanger from a residual heat discharging pipeline in the form of steam through a steam generator, and form condensate water after being cooled by a water tank to flow back into the steam generator; the water vapor formed by water heat exchange in the cooling water tank is discharged from the open type thermosiphon, partial water vapor is condensed during discharge and can flow back into the water tank, the beneficial effect of self-holding of water tank water quantity is realized, the consumption speed of water tank water quantity during nuclear power station accidents is greatly reduced, the cooling capacity is improved, the sustainable cooling time is prolonged, and the passive waste heat discharge requirement under the limited application environment of the floating nuclear power station space is met.
In the description of the present invention, it should be understood that the terms "longitudinal," "transverse," "upper," "lower," "front," "rear," "left," "right," "vertical," "horizontal," "top," "bottom," "inner," "outer," and the like indicate or are based on the orientation or positional relationship shown in the drawings, merely to facilitate description of the present invention, and do not indicate or imply that the devices or elements referred to must have a particular orientation, be constructed and operated in a particular orientation, and thus should not be construed as limiting the present invention.
The foregoing embodiments are merely illustrative of the preferred embodiments of the present invention, and the scope of the present invention is not limited thereto, but various modifications and improvements made by those skilled in the art to which the present invention pertains are made without departing from the spirit of the present invention, and all changes and modifications and improvements fall within the scope of the present invention as defined in the appended claims.
Claims (5)
1. A self-sustaining floating nuclear power plant passive waste heat removal system, the floating nuclear power plant comprising a steam generator (17), the steam generator (17) being in communication with a main steam line and a main feedwater line, comprising:
The heat pump comprises a C-shaped tube heat exchanger (1), wherein an inlet (7) of the C-shaped tube heat exchanger (1) is communicated with one end of a waste heat discharge pipeline (11), and the other end of the waste heat discharge pipeline (11) is communicated with the steam generator (17) or the main steam pipeline; an outlet (8) of the C-shaped tube heat exchanger (1) is communicated with one end of a condensed water pipeline (12), and the other end of the condensed water pipeline (12) is communicated with the steam generator (17) or the main water supply pipeline;
the cooling water tank (2), the C-shaped tube heat exchanger (1) is arranged in the cooling water tank (2);
The plurality of open thermosiphons (6) are arranged, the plurality of open thermosiphons (6) are fixedly arranged on the upper surface of the cooling water tank (2), and the bottom end of the open thermosiphons (6) penetrates through the upper surface of the cooling water tank (2) and is communicated with the cooling water tank (2); one or more of a silk screen, a groove and a steel powder sintered layer are arranged on the inner side wall of the open type thermosiphon (6);
a first isolation valve (9), wherein the first isolation valve (9) is arranged on the waste heat discharging pipeline (11);
a second isolation valve (13), the second isolation valve (13) being arranged on the condensate pipe (12);
the cooling tower (3), a plurality of open thermosiphons (6) set up in cooling tower (3) and be close to the bottom setting of cooling tower (3), cooling tower (3) bottom has cold air inlet (4), and the top has hot air outlet (5).
2. A self-sustaining floating nuclear power plant passive waste heat removal system as claimed in claim 1, wherein a third isolation valve (15) is provided on the main steam line and a fourth isolation valve (14) is provided on the main feedwater line.
3. A self-sustaining floating nuclear power plant passive residual heat removal system according to claim 2, characterized in that when the other end of the residual heat removal pipe (11) is in communication with the main steam pipe, the communication is between the third isolation valve (15) and the steam generator (17).
4. A self-sustaining floating nuclear power plant passive residual heat removal system according to claim 2, wherein when the other end of the condensate conduit (12) is in communication with the main feed water conduit, the communication is between the fourth isolation valve (14) and the steam generator (17).
5. A self-sustaining floating nuclear power plant passive residual heat removal system according to claim 1, further comprising a check valve (10), said check valve (10) being arranged on said condensate line (12) between said second isolation valve (13) and said steam generator (17).
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202310714834.8A CN117095840B (en) | 2023-06-16 | 2023-06-16 | Self-supporting passive waste heat discharging system of floating nuclear power station |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202310714834.8A CN117095840B (en) | 2023-06-16 | 2023-06-16 | Self-supporting passive waste heat discharging system of floating nuclear power station |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN117095840A CN117095840A (en) | 2023-11-21 |
| CN117095840B true CN117095840B (en) | 2024-05-10 |
Family
ID=88778089
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202310714834.8A Active CN117095840B (en) | 2023-06-16 | 2023-06-16 | Self-supporting passive waste heat discharging system of floating nuclear power station |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN117095840B (en) |
Citations (13)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN201946323U (en) * | 2011-01-05 | 2011-08-24 | 中科华核电技术研究院有限公司 | Emergency water supply system for nuclear power station |
| CN102867548A (en) * | 2012-09-27 | 2013-01-09 | 中国核电工程有限公司 | Active and passive combined secondary side reactor core heat derivation device |
| CN202887750U (en) * | 2012-09-27 | 2013-04-17 | 中国核电工程有限公司 | Advanced secondary side reactor core heat leading-out device |
| CN103617815A (en) * | 2013-12-05 | 2014-03-05 | 哈尔滨工程大学 | Passive residual heat exhausting system of pressurized water reactor nuclear power plant |
| CN204407021U (en) * | 2015-01-30 | 2015-06-17 | 中广核工程有限公司 | Nuclear power station non-active spentnuclear fuel pond emergence compensating water system |
| KR101576077B1 (en) * | 2014-07-23 | 2015-12-10 | 한국원자력연구원 | Passive residual heat removal system and nuclear power plant having the same |
| CN105957567A (en) * | 2016-05-06 | 2016-09-21 | 中国核动力研究设计院 | Steam generator secondary side passive residual heat removal system |
| KR101677981B1 (en) * | 2015-09-04 | 2016-11-21 | 한국원자력연구원 | Safty system for a nuclear power plant and nuclear power plant having the same |
| CN107403650A (en) * | 2017-08-25 | 2017-11-28 | 中国船舶重工集团公司第七〇九研究所 | The Passive residual heat removal system of floating nuclear power plant |
| RU2740786C1 (en) * | 2020-04-03 | 2021-01-21 | федеральное государственное бюджетное образовательное учреждение высшего образования "Санкт-Петербургский государственный морской технический университет" (СПбГМТУ) | System for passive heat removal of reactor plant |
| KR20210118626A (en) * | 2020-03-23 | 2021-10-01 | 한국수력원자력 주식회사 | Passive residual heat removal system |
| KR20210158164A (en) * | 2020-06-23 | 2021-12-30 | 한국수력원자력 주식회사 | Passive residual heat removal system of integral reactor for floating structure |
| CN115240880A (en) * | 2022-08-01 | 2022-10-25 | 哈尔滨工程大学 | Passive residual heat removal system and method capable of achieving continuous heat removal |
-
2023
- 2023-06-16 CN CN202310714834.8A patent/CN117095840B/en active Active
Patent Citations (13)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN201946323U (en) * | 2011-01-05 | 2011-08-24 | 中科华核电技术研究院有限公司 | Emergency water supply system for nuclear power station |
| CN102867548A (en) * | 2012-09-27 | 2013-01-09 | 中国核电工程有限公司 | Active and passive combined secondary side reactor core heat derivation device |
| CN202887750U (en) * | 2012-09-27 | 2013-04-17 | 中国核电工程有限公司 | Advanced secondary side reactor core heat leading-out device |
| CN103617815A (en) * | 2013-12-05 | 2014-03-05 | 哈尔滨工程大学 | Passive residual heat exhausting system of pressurized water reactor nuclear power plant |
| KR101576077B1 (en) * | 2014-07-23 | 2015-12-10 | 한국원자력연구원 | Passive residual heat removal system and nuclear power plant having the same |
| CN204407021U (en) * | 2015-01-30 | 2015-06-17 | 中广核工程有限公司 | Nuclear power station non-active spentnuclear fuel pond emergence compensating water system |
| KR101677981B1 (en) * | 2015-09-04 | 2016-11-21 | 한국원자력연구원 | Safty system for a nuclear power plant and nuclear power plant having the same |
| CN105957567A (en) * | 2016-05-06 | 2016-09-21 | 中国核动力研究设计院 | Steam generator secondary side passive residual heat removal system |
| CN107403650A (en) * | 2017-08-25 | 2017-11-28 | 中国船舶重工集团公司第七〇九研究所 | The Passive residual heat removal system of floating nuclear power plant |
| KR20210118626A (en) * | 2020-03-23 | 2021-10-01 | 한국수력원자력 주식회사 | Passive residual heat removal system |
| RU2740786C1 (en) * | 2020-04-03 | 2021-01-21 | федеральное государственное бюджетное образовательное учреждение высшего образования "Санкт-Петербургский государственный морской технический университет" (СПбГМТУ) | System for passive heat removal of reactor plant |
| KR20210158164A (en) * | 2020-06-23 | 2021-12-30 | 한국수력원자력 주식회사 | Passive residual heat removal system of integral reactor for floating structure |
| CN115240880A (en) * | 2022-08-01 | 2022-10-25 | 哈尔滨工程大学 | Passive residual heat removal system and method capable of achieving continuous heat removal |
Non-Patent Citations (4)
| Title |
|---|
| 二次侧非能动余热排出系统设计及验证分析;陈薇;张亚男;曹夏昕;阎昌琪;张往锁;;中国核电;20120615(第02期);全文 * |
| 代守宝 ; 彭敏俊 ; .IPWRs非能动余热排出系统运行特性分析.核科学与工程.2010,(第03期),全文. * |
| 耿一娲 ; .非能动安全系统在AP1000核电站应用概述.科学中国人.2017,(第15期),全文. * |
| 非能动余热排出换热器在主给水管道断裂事故下的冷却能力研究;肖三平;陈树山;吴昊;;原子能科学技术;20160320(第03期);全文 * |
Also Published As
| Publication number | Publication date |
|---|---|
| CN117095840A (en) | 2023-11-21 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN103377728B (en) | A kind of Flooded-type containment complete passive after-heat removal system | |
| CN106816186A (en) | A kind of Integrated PWR Heat Discharging System of Chinese based on separate heat pipe | |
| CN107393605A (en) | The passive air-cooling apparatus and method of a kind of modular small nuclear reactor | |
| JP2017521674A (en) | PCV cooling system, and PCV / reactor pressure vessel joint cooling system | |
| CN104916334A (en) | Separated heat pipe type passive residual heat removal system for pressurized water reactor nuclear power plant | |
| CN107464590A (en) | Marine PWR Passive residual heat removal system | |
| CN107644693B (en) | Naval reactor and once through steam generator Passive residual heat removal system | |
| CN103617815A (en) | Passive residual heat exhausting system of pressurized water reactor nuclear power plant | |
| CN104913364B (en) | A kind of solar energy soil source hot water type absorption heat pump heating system | |
| CN107492400B (en) | Dry reactor heating system | |
| CN107403650A (en) | The Passive residual heat removal system of floating nuclear power plant | |
| CN204455305U (en) | A kind of containing tritium heavy water self-circulation electrolytic system | |
| CN107170493A (en) | A kind of passive containment thermal conduction system | |
| US20230223160A1 (en) | Reactor secondary side passive residual heat removal system | |
| KR101505475B1 (en) | Passive containment cooling system and nuclear power plant having the same | |
| CN205656860U (en) | Active discharge system of reactor core waste heat non - is piled in heat supply of low temperature nuclear | |
| CN105118534B (en) | Passive Spent Fuel Pool cooling and water charging system | |
| CN105976873A (en) | Internal part cooling power generation system for future tokamak fusion reactor | |
| CN210271804U (en) | Emergent waste heat discharge system of water injection formula lead bismuth fast reactor | |
| CN204680390U (en) | Pressurized-water reactor nuclear power plant separate heat pipe formula Heat Discharging System of Chinese | |
| CN207624389U (en) | Heat exchanger is located at the outer marine PWR Passive residual heat removal system of water tank | |
| CN117095840B (en) | Self-supporting passive waste heat discharging system of floating nuclear power station | |
| CN113035387A (en) | PCS (Power distribution System) long-term cooling water tank capable of operating efficiently | |
| CN115240880B (en) | Passive residual heat removal system and method capable of achieving continuous heat removal | |
| CN207250149U (en) | The Passive residual heat removal system of floating nuclear power plant |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| PB01 | Publication | ||
| PB01 | Publication | ||
| SE01 | Entry into force of request for substantive examination | ||
| SE01 | Entry into force of request for substantive examination | ||
| GR01 | Patent grant | ||
| GR01 | Patent grant |