CN204614459U - A kind of non-active nuclear power station pressure release condensation heat exchange system - Google Patents
A kind of non-active nuclear power station pressure release condensation heat exchange system Download PDFInfo
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- CN204614459U CN204614459U CN201420847850.0U CN201420847850U CN204614459U CN 204614459 U CN204614459 U CN 204614459U CN 201420847850 U CN201420847850 U CN 201420847850U CN 204614459 U CN204614459 U CN 204614459U
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Abstract
A kind of non-active nuclear power station pressure release condensation heat exchange system, it comprises the steam header and closed natural convection loop that are communicated with main automatic dropping valve, closed natural convection loop comprises non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, the outer non-active heat exchanger of shell and heat transferring medium, wherein steam header is arranged in containment, the top of steam header is configured with discharge of steam pipeline, the bottom of steam header is configured with condensed water elimination pipeline, heat-exchanging loop pipeline runs through steam header and containment is arranged, non-active vapor condensation heat-exchange device is arranged in steam header and is communicated with heat-exchanging loop pipeline, it is outer and be communicated with heat-exchanging loop pipeline that the outer non-active heat exchanger of shell is arranged on containment, heat transferring medium is absorbed heat at non-active vapor condensation heat-exchange device and is transferred heat to the outer non-active heat exchanger of shell by heat-exchanging loop pipeline, thus establish closed Natural Circulation, to continue taking away reactor core remnants fission producing residual heat of nuclear core when having an accident.
Description
Technical field
The utility model relates to a kind of heat-exchange system, in particular to the non-active nuclear power station pressure release condensation heat exchange system of one.
Background technology
The nuclear power of safety is a kind of clean energy resource of high-energy source density, to preserving the ecological environment, readjusting the energy structure and ensure that energy security has important effect.But once safety problem appears in nuclear power station, then can bring huge threat to staff, nearby residents and ecologic environment etc.Nuclear plant safety problem is that people must the problem considered of emphasis when applying nuclear power for this reason.Current nuclear power station tends to adopt non-passive safety technical finesse accident.So-called non-passive safety technology refers to and utilizes natural force to complete various refrigerating function in the situation of having an accident, and the driving force etc. that wherein natural force can be produced by gravity, pressure accumulation gaseous tension, Natural Circulation produce, without the need to pump and external power source.Therefore, while improve nuclear plant safety reliability, enormously simplify the security system of nuclear power station.
The reactor core cooling system (will describe in detail in specific embodiment part) that the non-active nuclear power station of prior art comprises primary heat transport system and is communicated with it, reactor core cooling system is used for taking away when having an accident the reactor core waste heat that in primary heat transport system, reactor core remnants fission produces.
It is emphasized that in the prior art, main automatic dropping valve has maximum circulation discharge area, and the steam that in the primary heat transport system of being discharged in containment by it, residual heat of nuclear core produces accounts for major portion.This will cause following problem: 1) in containment, pressure will increase, affect the pressure releasing of primary heat transport system conversely, postpone the startup of main water supply tank to the gravity water filling of reactor pressure vessel, and now the first water supply tank and the second water supply tank substantially emptying, reactor core is in liquid level in accident minimum dangerous period, and pressure release now delays to add the exposed risk of reactor core.For the test of prior art and analytical proof (see what within 1999, to deliver at " Nuclear Engineering and Design ", author is David E. Bessette, Marino di Marzo's and name be called the document of " Transition from depressurization to long term cooling in AP600 scaled integral test facilities "), reactor core minimum liquid level will be there is in this transition period that main automatic dropping valve is opened between the startup of main water supply tank peace note, therefore for existing non-passive safety technology, primary heat transport system pressure reduction after main automatic dropping valve unlatching is the distress phase paid close attention to.2) because the containment cooling system of non-active nuclear power station also adopts non-enabling fashion, after the outer chilled water of containment drains, (chilled water flow by gravity as outer in existing AP1000 technology shell can maintain 72 hours, emergently outside factory after supposing 72 hours can provide electrical source of power and recover the outer cooling water of shell) long-term natural cooling stage, the heat exchange mode of heat transfer and natural convection air such as can only be relied on to carry out condensation to the steam in containment, and containment keeps sufficient exchange capability of heat to face the challenge for a long time.3) steam condensing reflux is spurted in the pressure release through leading automatic dropping valve needs a period of time to melt pit, especially cooling deficiency outward at containment causes steam-condensation in shell to there will be the reduction of melt pit liquid level not in time, and the natural circulation cooling flow of reactor core is reduced.These all will cause reactor core to cool risk.
The utility model aims to provide a kind of novel non-active nuclear power station pressure release condensation heat exchange system and is connected with major loop hot arc through main automatic dropping valve, make to spurt through main automatic dropping valve the steam that and heat is no longer assembled in containment, can effectively reduce pressure in containment, promote that main automatic dropping valve is primary heat transport system pressure release, guarantee that main water supply tank water filling starts in time, alleviate the situation that the cooling of this distress phase reactor core is not enough; Improve the limitation that post incident relies on containment long term air to cool completely simultaneously; If by steam rapid condensation and the melt pit that refluxes, can effectively keep melt pit to flood liquid level, maintain stable reactor core natural circulation cooling flow, promote the security of long-term cooling; Thus the deficiency that the passive safety system overcoming current non-active nuclear power station exists.
Utility model content
An embodiment of the present utility model provides a kind of non-active nuclear power station pressure release condensation heat exchange system, wherein non-active nuclear power station comprises containment and main automatic dropping valve, main automatic dropping valve is communicated with the major loop hot arc be connected on reactor pressure vessel, for the steam in the release reaction core pressure vessel when having an accident, wherein non-active nuclear power station pressure release condensation heat exchange system comprises the steam header and closed natural convection loop that arrange and be communicated with main automatic dropping valve, closed natural convection loop comprises non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, the outer non-active heat exchanger of shell and heat transferring medium, wherein steam header is arranged in the containment of non-active nuclear power station, the top of steam header is configured with steam header discharge of steam pipeline, the bottom of steam header is configured with the steam header condensed water elimination pipeline being provided with retaining valve, heat-exchanging loop pipeline runs through steam header and containment is arranged, non-active vapor condensation heat-exchange device to be arranged in steam header and to be communicated with heat-exchanging loop pipeline, it is outer and be communicated with heat-exchanging loop pipeline that the outer non-active heat exchanger of shell is arranged on containment, the outer non-active heat exchanger of shell is arranged on higher position relative to non-active vapor condensation heat-exchange device, heat transferring medium in closed natural convection loop is absorbed heat at non-active vapor condensation heat-exchange device and is transferred heat to the outer non-active heat exchanger of shell by heat-exchanging loop pipeline, thus at non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, closed Natural Circulation is established between the outer non-active heat exchanger of shell, to continue taking away reactor core remnants fission producing reactor core waste heat when non-active nuclear power station has an accident.
According to the non-active nuclear power station pressure release condensation heat exchange system that an embodiment of the present utility model provides, wherein closed natural convection loop adopts hot tube heat exchanger principle, its heat transfer medium can be the potpourri of cooling water and steam, closed natural convection loop is evacuated, when main automatic dropping valve is opened and when the steam in reactor pressure vessel is discharged into steam header, main automatic dropping valve spurts the steam and is discharged in steam header, main automatic dropping valve spurts the steam that through non-active vapor condensation heat-exchange device, heat transfer medium in non-active vapor condensation heat-exchange device is heated and forms condensate water by heat exchange, condensate water is discharged in melt pit by the steam header condensed water elimination pipeline being configured in the bottom of steam header, thus supplement chilled water to melt pit, ensure the stability of the long-term cool cycles of reactor core, chilled water in closed natural convection loop when main automatic dropping valve is opened and its temperature exceedes the design temperature of the closed Natural Circulation starting closed natural convection loop time added thermosetting steam at non-active vapor condensation heat-exchange device place, steam in closed natural convection loop is outer non-active heat exchanger flowing along heat-exchanging loop pipeline towards shell, heat is discharged into the atmosphere by the outer non-active heat exchanger of shell, steam in closed natural convection loop forms condensate water after cooling and relies on gravity again to get back to non-active vapor condensation heat-exchange device in the outer non-active heat exchanger of shell, thus closed Natural Circulation is established in closed natural convection loop, to continue the reactor core waste heat of the reactor core remnants fission generation taken away in reactor pressure vessel when having an accident.
According to the non-active nuclear power station pressure release condensation heat exchange system that another embodiment of the present utility model provides, wherein closed natural convection loop also comprises the outer heat exchange isolation valve of shell, the outer heat exchange isolation valve of shell is arranged between non-active vapor condensation heat-exchange device and the outer non-active heat exchanger of shell along heat transferring medium flowing direction in closed natural convection loop, and the outer heat exchange isolation valve of shell is interlocked with main automatic dropping valve and opened; when main automatic dropping valve is opened and when the steam in reactor pressure vessel is discharged into steam header, steam is through non-active vapor condensation heat-exchange device, heat transferring medium in non-active vapor condensation heat-exchange device is heated and forms condensate water by heat exchange, condensate water is discharged in melt pit by the steam header condensed water elimination pipeline and the retaining valve be arranged on steam header condensed water elimination pipeline being configured in the bottom of steam header, flowed towards the outer non-active heat exchanger of shell along closed natural convection loop by the heat transferring medium heated in non-active vapor condensation heat-exchange device, heat is discharged into the atmosphere by the outer non-active heat exchanger of shell, in the outer non-active heat exchanger of shell, cooled heat transferring medium relies on gravity again to get back to non-active vapor condensation heat-exchange device, thus at non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, closed Natural Circulation is established between the outer non-active heat exchanger of shell and the outer heat exchange isolation valve of shell, to continue the reactor core waste heat of the reactor core remnants fission generation taken away in reactor pressure vessel when having an accident.
According to the non-active nuclear power station pressure release condensation heat exchange system that above-mentioned embodiment of the present utility model provides, wherein can not drain in containment via the discharge of steam pipeline at steam header top by the fouling gas that contains of the steam in the steam of total condensation or steam header in steam header, the heat in containment be discharged into the atmosphere by containment cooling system.
According to the non-active nuclear power station pressure release condensation heat exchange system that above-mentioned embodiment of the present utility model provides, the wherein non-active nuclear power station reactor core cooling system that comprises primary heat transport system and be communicated with it, reactor core cooling system is used for taking away when having an accident the reactor core waste heat that in primary heat transport system, reactor core remnants fission produces.
According to the non-active nuclear power station pressure release condensation heat exchange system that above-mentioned embodiment of the present utility model provides, wherein primary heat transport system comprises steam generator, U-tube, cold section of major loop, major loop hot arc, main pump, reactor pressure vessel, be positioned at the reactor core of reactor pressure vessel, Surge line piping and voltage stabilizer, wherein U-tube is arranged in a vapor generator, U-tube endpiece is communicated with through cold section of main pump and major loop through the cold chamber compartment by steam generator bottom, cold section of major loop is communicated with reactor pressure vessel, reactor pressure vessel is communicated with major loop hot arc, major loop hot arc to be communicated with voltage stabilizer by Surge line piping and the hot chamber compartment passing through steam generator bottom is communicated with the inlet end of U-tube, cooling medium enters reactor pressure vessel by cold section of major loop, arrive the entrance of reactor core, the Q-value that reactor core produces is taken away when flowing through reactor core, major loop hot arc is flowed through by the cooling medium heated, arrive the hot chamber compartment of steam generator bottom and enter the inlet end of U-tube, to be transferred heat in steam generator by U-tube and cooling medium outside U-tube, coolant temperature in U-tube reduces and collects in the cold chamber compartment of steam generator bottom by the endpiece of U-tube, the main pump that cooling medium in cold chamber compartment is communicated with cold cavity bottom pumps into cold section of major loop, again get back to reactor pressure vessel, form the enclosed cool cycles of primary heat transport system.
According to the non-active nuclear power station pressure release condensation heat exchange system that above-mentioned embodiment of the present utility model provides, wherein reactor core cooling system comprises the first water supply tank, second water supply tank, main water supply tank, be arranged in the passive residual heat removal heat exchanger of main water supply tank, level Four Automatic Depressurization System, melt pit, melt pit filter screen, melt pit return line and the explosive valve be arranged on melt pit return line, first water supply tank, second water supply tank, main water supply tank is communicated with reactor pressure vessel by direct reactor peace note pipe respectively by corresponding connecting line and the retaining valve be arranged on each connecting line, first water supply tank top is communicated with by cold section of pressure-equalizing line and major loop, thus the pressure of the pressure in the first water supply tank and primary heat transport system is consistent, level Four Automatic Depressurization System comprises the automatic dropping valve of the first order, the automatic dropping valve in the second level, the automatic dropping valve of the third level and main automatic dropping valve, the automatic dropping valve of the first order, the automatic dropping valve in the second level, the inlet end of the automatic dropping valve of the third level to be connected on voltage stabilizer and the automatic dropping valve of the first order with parallel way, the automatic dropping valve in the second level, the endpiece of the automatic dropping valve of the third level is connected on main water supply tank with parallel way, main automatic dropping valve is communicated with major loop hot arc, passive residual heat removal heat exchanger is communicated with major loop hot arc with cold section of major loop, Natural Circulation is established at passive residual heat removal heat exchanger and between cold section of major loop and major loop hot arc, reactor pressure vessel is arranged in melt pit, chilled water in melt pit is by melt pit return line, melt pit filter screen be arranged on melt pit return line borehole blasting valve and be communicated with reactor pressure vessel by direct safety injection pipe.
Have the following advantages according to non-active nuclear power station pressure release condensation heat exchange system of the present utility model: 1) can by outside reactor core Residual heat removal containment, the pressurized effect that in effective reduction containment, steam is assembled, decrease the load of containment cooling system, the chilled water gravity flow outside shell being conducive to realizing containment relies on cross-ventilated long-term cooling after draining simultaneously.2) decrease discharge of steam in containment, reduce the back pressure of main automatic dropping valve discharge, be conducive to accelerating to reduce the pressure in pressure vessel, guarantee that main water supply tank drops into and sustained water injection in time, thus make reactor core be in safer flooding and the state of cooling.3) adopt non-active vapor condensation heat-exchange device to carry out condensation to the steam that main automatic dropping valve gives off and become condensate water, condensate water returns melt pit by the condensed water elimination pipeline bottom steam tank, supplement chilled water to melt pit, guarantee that long-term melt pit recycle cooling is carried out sustainedly and stably.4) the non-active nuclear power station pressure release condensation heat exchange system of setting up and reactor core medium completely isolated, decrease the risk that radioactivity leaks outside.5) do not change existing non-active core cooling system structure, and non-active nuclear power station pressure release condensation heat exchange system of the present utility model adopts non-enabling fashion, rely on natural force to drive, keep original non-active design concept.
Accompanying drawing explanation
Above and other aspect of the present utility model is discussed in detail, in accompanying drawing below in conjunction with accompanying drawing:
Fig. 1 is the non-active PIPING OF MAIN LOOPS IN NUCLEAR POWER STATION system schematic of prior art.
Fig. 2 is the sketch of the non-active nuclear power station reactor core cooling system of prior art.
Fig. 3 is the sketch of the long-term cool cycles process of melt pit of the non-active nuclear power station reactor core cooling system of prior art.
Fig. 4 is the schematic diagram of the non-active pressure release condensation heat exchange system according to an embodiment of the present utility model.
Fig. 5 is the schematic diagram of the non-active pressure release condensation heat exchange system according to another embodiment of the present utility model.
Parts and label list
| 1 | Reactor core |
| 2 | Reactor pressure vessel |
| 3 | Cold section of major loop |
| 4 | Major loop hot arc |
| 5 | U-tube |
| 6 | Steam generator |
| 7 | The cold chamber compartment of steam generator |
| 8 | The hot chamber compartment of steam generator |
| 9 | Main pump |
| 10 | Surge line piping |
| 11 | Voltage stabilizer |
| 12 | Main steam pipe |
| 13 | Main steam isolation valve |
| 14 | Passive residual heat removal heat exchanger |
| 15 | First water supply tank |
| 16 | Second water supply tank |
| 17 | Main water supply tank |
| 18 | Direct safety injection pipe |
| 19 | Pressure-equalizing line |
| 20 | The automatic dropping valve of the first order |
| 21 | The automatic dropping valve in the second level |
| 22 | The automatic dropping valve of the third level |
| 23 | Main automatic dropping valve |
| 24 | Melt pit filter screen |
| 25 | Containment |
| 26 | Steam header |
| 27 | Non-active vapor condensation heat-exchange device |
| 30 | The outer non-active heat interchanger of shell |
| 31 | Heat-exchanging loop pipeline |
| 32 | The outer heat exchange isolation valve of shell |
| 29，51-53 | Retaining valve |
| 54-55 | Explosive valve |
| 56 | Stop valve |
| 60 | Bubbler |
| 100 | First connecting line |
| 102 | Second connecting line |
| 104 | 3rd connecting line |
| 105 | Melt pit |
| 106 | Melt pit return line |
| 108 | Steam header discharge of steam pipeline |
| 110 | Steam header condensed water elimination pipeline |
Embodiment
Fig. 1-Fig. 5 and following description describe Alternate embodiments of the present utility model and how to implement to instruct those of ordinary skill in the art and to reproduce the utility model.In order to instruct technical solutions of the utility model, simplifying or having eliminated some conventional aspects.It should be understood by one skilled in the art that the modification that is derived from these embodiments or replace and will drop in protection domain of the present utility model.It should be understood by one skilled in the art that following characteristics can combine to form multiple modification of the present utility model in every way.Thus, the utility model is not limited to following Alternate embodiments, and only by claim and their equivalents.
Cooling medium can be such as chilled water in this article.Other cooling mediums being adapted at nuclear power plant system use of refrigerating function can be realized also in scope of the present utility model.
Fig. 1 shows the primary heat transport system of current non-active nuclear power station.As shown in Figure 1, the primary heat transport system of current non-active nuclear power station comprises steam generator 6, U-tube 5, major loop cold section 3, major loop hot arc 4, main pump 9, reactor pressure vessel 2, be positioned at the reactor core 1 of reactor pressure vessel 2, Surge line piping 10 and voltage stabilizer 11, wherein U-tube 5 is arranged in steam generator 6, U-tube endpiece is pooled to the cold chamber compartment 7 of steam generator bottom, cold chamber compartment 7 is communicated with for cold section 3 with major loop by main pump 9, major loop is communicated with reactor pressure vessel 2 for cold section 3, reactor pressure vessel 2 is also communicated with major loop hot arc 4, major loop hot arc 4 to be communicated with voltage stabilizer 11 by Surge line piping 10 and to be communicated with the inlet end of U-tube 5 by the hot chamber compartment 8 of steam generator bottom, cooling medium enters reactor pressure vessel 2 for cold section 3 by major loop, arrive the entrance of reactor core 1, the Q-value that reactor core produces is taken away when flowing through reactor core 1, major loop hot arc 4 is flowed through by the cooling medium (such as temperature is about 321 DEG C) heated, arrive the hot chamber compartment 8 of steam generator bottom and enter the inlet end of U-tube 5, the cooling medium in steam generator 6 and outside U-tube 5 is transferred heat to by U-tube 5, coolant temperature in U-tube 5 reduces (such as coolant temperature is 280 DEG C) and collects in the cold chamber compartment 7 of steam generator bottom by the endpiece of U-tube, in cold chamber compartment, the cooling medium of 7 pumps into major loop cold section 3 by main pump 9, again get back to reactor pressure vessel 2, form the enclosed cool cycles of primary heat transport system.In Fig. 1, arrow F1 is that the cooling medium that temperature is lower flows to, and arrow F2 is that the cooling medium that temperature is higher flows to.
In order to the pressure of stable primary heat transport system, major loop hot arc 4 is communicated with voltage stabilizer 11 by Surge line piping 10, is saturated solution and saturated vapour (being such as saturated solution and the saturated vapour of chilled water), meets the voltage stabilizing requirement of primary heat transport system in voltage stabilizer 11.Voltage stabilizer 11 is for maintaining the normal high pressure conditions (as about 15.5MPa) run by the pressure of primary heat transport system, make between reactor core 1 reaction period in normal operation, the cooling medium in reactor pressure vessel 2 there will not be boiling.Through reactor core 1 heat chilled water when flowing through U-tube 5, transfer heat to the chilled water in steam generator 6 and outside U-tube 5, the cooling water evaporation in steam generator 6 is made to form steam, in steam generator 6, steam is by main steam pipe 12, steam turbine (not shown in figure 1) is fed to by the main steam isolation valve 13 often opened, drive steam turbine generates electricity, thus the heat energy produced by reactor core is electric energy.
But; when there is minor break accident in primary heat transport system; although reactor core stops reaction; but reactor core remnants fission is still in continuation; still produce a large amount of waste heat (being such as equivalent to the 1%-6% of normal power); main pump 9 due to now primary heat transport system shuts down main steam isolation valve 13 simultaneously and cuts out, and cannot normally take away reactor core waste heat.Thus, if do not start non-active nuclear power station reactor core cooling system, then reactor core will melt and develop into serious accident by overtemperature.
Fig. 2 is the sketch of the non-active nuclear power station reactor core cooling system of prior art.Such as, be that Ouyang gives see 2010 authors published by Atomic Energy Press, Lin Chengge's etc. and name be called the document of " non-passive safety advanced pressurized water reactor nuclear power technology ".As shown in Figure 2, the reactor core cooling system of prior art comprises the single chilled water storing about 70 tons of the first water supply tank 15(), the second water supply tank 16(is single stores the chilled water of about 57 tons and the gas of about 5MPa), main water supply tank 17(stores the water of about 2100 tons), the explosive valve 55 that is arranged in the passive residual heat removal heat exchanger 14 of main water supply tank 17 and bubbler 60, level Four Automatic Depressurization System, melt pit 105, melt pit filter screen 24, melt pit return line 106 and is arranged on melt pit return line.First water supply tank 15, second water supply tank 16, main water supply tank 17 are respectively by the first connecting line 100, second connecting line 102, the 3rd connecting line 104 and the retaining valve 51-53 that is arranged on each connecting line and be communicated with reactor pressure vessel 2 by direct reactor peace note pipe 18, wherein the first connecting line 100 can be provided with stop valve 56,3rd connecting line 104 can be provided with explosive valve 54, and stop valve 56 and explosive valve 54 are all for preventing the improper injection of water tank inner cooling water.First water supply tank top is communicated with for cold section 3 with major loop by pressure-equalizing line 19, thus the pressure of the pressure in the first water supply tank 15 and primary heat transport system is consistent, level Four Automatic Depressurization System comprises the automatic dropping valve 20 of the first order, the automatic dropping valve 21 in the second level, the automatic dropping valve of the third level 22 and main automatic dropping valve 23, the automatic dropping valve 20 of the first order, the automatic dropping valve 21 in the second level, the inlet end of the automatic dropping valve of the third level 22 to be connected on voltage stabilizer 11 with parallel way and by automatic for first order dropping valve 20, the automatic dropping valve 21 in the second level, the endpiece of the automatic dropping valve of the third level 22 is connected on the bubbler 60 that is positioned in main water supply tank 17 with parallel way, main automatic dropping valve 23 is communicated with major loop hot arc 4, cold section of 3(is not shown for passive residual heat removal heat exchanger 14 and major loop) be communicated with major loop hot arc 4, coolant density difference is relied on to establish Natural Circulation at passive residual heat removal heat exchanger 14 and between major loop cold section 3 and major loop hot arc 4.Reactor pressure vessel 2 is arranged in melt pit 105, and the chilled water in melt pit 105 is by melt pit return line 106, melt pit filter screen 24 and the explosive valve 55 that is arranged in melt pit return line and be communicated with reactor pressure vessel 2 by direct safety injection pipe 18.
Usual minor break accident process comprises five typical stages: 1) underheat blowdown phase; 2) the saturated Natural Circulation stage; 3) dropping valve triggers buck stage automatically; 4) the main water supply tank gravity safety injection stage; And 5) long-term melt pit recycle cooling stage.
At underheat blowdown phase, the chilled water in primary heat transport system spurts in containment from cut, the pressure drop in reactor pressure vessel 2.Chilled water in primary heat transport system reduces, and causes the water level in voltage stabilizer 11 to reduce, will trigger shutdown and safety signal, and reactor core 1 stops reaction, and main pump 9 is shut down, and the main steam isolation valve 13 on the main steam pipe 12 that steam generator 6 exports is closed.Stop valve 56 is opened simultaneously, and passive residual heat removal heat exchanger 14 and the first water supply tank 15 rely on natural driving force to put into operation.
In the saturated blowdown Natural Circulation stage, the Pressure Drop in reactor pressure vessel 2 is low to moderate primary heat transport system saturation pressure (as being 7.6MPa), occurs Gas-liquid two-phase flow in reactor pressure vessel 2.Between passive residual heat removal heat exchanger 14, major loop hot arc 4 and major loop cold section 3, rely on coolant density difference to establish Natural Circulation.Meanwhile, the first water supply tank 15 relies on gravity via the first associated line 100 and retaining valve 51 disposed thereon and in reactor pressure vessel 2, injects chilled water by direct safety injection pipeline 18.
Automatic step-down triggers buck stage and refers to when water level in the first water supply tank 15 is reduced to the first level set value (as corresponded to the liquid level of 75% volume in the first water supply tank 15), the automatic dropping valve 20 of the first order will be triggered, after the time delay of setting, open the automatic dropping valve in the second level 21 and the automatic dropping valve 22 of the third level successively.The triggering of the automatic dropping valve of the first order 20, the automatic dropping valve in the second level 21 and the automatic dropping valve of the third level 22 causes the pressure in reactor pressure vessel 2 to accelerate to decline (as rapidly dropped to about 0.5MPa from about 7.6MPa), when primary heat transport system pressure is lower than (such as the about 5MPa of pressure accumulation nitrogen pressure) during pressure accumulation gaseous tension in the second water supply tank 16, the chilled water in the second water supply tank 16 injects chilled water by direct safety injection pipeline 18 to reactor pressure vessel through the second associated line 102 and retaining valve 52 disposed thereon under gaseous tension drives.After the second water supply tank 16 is emptying, the first water supply tank 15 continues to inject chilled water to reactor pressure vessel.When water level decreasing to the second level set value (as corresponded to the liquid level of 25% volume in the first water supply tank 15) in the first water supply tank 15, by main for triggering automatic dropping valve 23, by the pressure reduced further in reactor pressure vessel 2 (being less than about 0.1MPa as being reduced to further from about 0.5MPa).
The main water supply tank 17 gravity safety injection stage refers to when the pressure in reactor pressure vessel 2 is discharged by main automatic dropping valve 23 and is reduced to close to atmospheric pressure, chilled water (have about 10m liquid level, correspond to about 0.1MPa pressure head) in main water supply tank 17 relies on gravity via the 3rd connecting line 104 and is arranged on retaining valve 53 wherein and injects chilled water by direct safety injection pipeline 18 to reactor pressure vessel 2.When the water level in main water supply tank is reduced to three-tank first level set value (such as the overall height liquid level of main water supply tank 40%), the explosive valve 55 that main water supply tank 17 will be opened before melt pit filter screen 24, main water supply tank 17 utilizes gravity to discharge water and cleans melt pit filter screen, prevents imminent melt pit from refluxing and blocks at melt pit filter screen place.When liquid level in melt pit to raise due to UNICOM's water filling of main water supply tank 17 reach melt pit liquid level concordant with the liquid level of main water supply tank 17 time, the cooling of melt pit long-term recirculation will be set up.
Fig. 3 is the sketch of the long-term cool cycles process of melt pit of the non-active nuclear power station reactor core cooling system of prior art.Can be described as: by melt pit filter screen 24, melt pit reflux pipeline 106 with in being arranged on melt pit reflux pipeline explosive valve 55 piii reactor pressure vessel 2 under the effect of the driving force that the chilled water collected in melt pit is formed at density difference, more than reactor core, thermogenetic steam is disposed in containment through main automatic dropping valve 23, establish Natural Circulation, and the steam in containment is cooled by containment cooling system, condensate water is back to melt pit in containment, supplements chilled water to melt pit.By such recycle design, reactor core waste heat is passed to as the ambient atmosphere outside the containment of ultimate heat sink, keep reactor core to continue cooling, prevent reactor core overtemperature from melting and developing even more serious accident.
The utility model aims to provide a kind of non-active nuclear power station pressure release condensation heat exchange system, to overcome the deficiency of non-active nuclear power station reactor core cooling system in prior art.
Fig. 1 is the PIPING OF MAIN LOOPS IN NUCLEAR POWER STATION system schematic of prior art.Fig. 4 is the schematic diagram of the non-active pressure release condensation heat exchange system according to an embodiment of the present utility model.As shown in Figure 1 and Figure 4, non-active nuclear power station comprises containment 25 and main automatic dropping valve 23, and main automatic dropping valve 23 is communicated with the major loop hot arc 4 be connected on reactor pressure vessel, for the steam in the release reaction core pressure vessel when having an accident.
Fig. 4 is the schematic diagram of the non-active pressure release condensation heat exchange system according to an embodiment of the present utility model.As shown in Figure 4, non-active nuclear power station pressure release condensation heat exchange system comprises the steam header 26 and closed natural convection loop that are communicated with main automatic dropping valve 23, closed natural convection loop comprises non-active vapor condensation heat-exchange device 27, heat-exchanging loop pipeline 31, the outer non-active heat exchanger 30 of shell, the outer heat exchange isolation valve 32 of shell and heat transfer medium, wherein steam header 26 is arranged in the containment 25 of non-active nuclear power station, the top of steam header is configured with steam header discharge of steam pipeline 108, the bottom of steam header is configured with the steam header condensed water elimination pipeline 110 being provided with retaining valve 29, heat-exchanging loop pipeline 31 runs through steam header 26 and containment 25 is arranged, non-active vapor condensation heat-exchange device 27 to be arranged in steam header 26 and to be communicated with heat-exchanging loop pipeline 31, it is outer and be communicated with heat-exchanging loop pipeline 31 that the outer non-active heat exchanger 30 of shell is arranged on containment 25, the outer non-active heat exchanger 30 of shell is arranged on higher position relative to non-active vapor condensation heat-exchange device 27, the outer heat exchange isolation valve 32 of shell is arranged between non-active vapor condensation heat-exchange device 27 and the outer non-active heat exchanger 30 of shell along heat transfer medium in closed natural convection loop (being such as chilled water) flow direction, the outer heat exchange isolation valve 32 of shell is interlocked with main automatic dropping valve 23 and is opened.
When main automatic dropping valve 23 is opened and the steam in reactor pressure vessel 2 is discharged into steam header 26, main automatic dropping valve 23 spurts the steam and is discharged in steam header 26 along the direction pointed by arrow F3, steam is through non-active vapor condensation heat-exchange device 27, heat transfer medium in non-active vapor condensation heat-exchange device is heated and forms condensate water by heat exchange, condensate water is discharged in melt pit 105 by the steam header condensed water elimination pipeline 110 and the retaining valve 29 be arranged on steam header condensed water elimination pipeline being configured in the bottom of steam header, thus supplement chilled water to melt pit 105, ensure the stability of the long-term cool cycles of reactor core.
In non-active vapor condensation heat-exchange device 27, by the heat transfer medium that heats, along heat-exchanging loop pipeline 31 towards shell, outer non-active heat exchanger 30 flows, heat is discharged into the atmosphere by the outer non-active heat exchanger 30 of shell, in the outer non-active heat exchanger of shell, cooled heat transfer medium relies on gravity again to get back to non-active vapor condensation heat-exchange device 27, thus at non-active vapor condensation heat-exchange device 27, heat-exchanging loop pipeline 31, the outer non-active heat exchanger 30 of shell, closed Natural Circulation is established between the outer heat exchange isolation valve 32 of shell, to continue the reactor core waste heat of the reactor core remnants fission generation taken away in reactor pressure vessel when having an accident.
Can not drain in containment via the discharge of steam pipeline 108 at steam header top by the gas of incoagulability that contains of the steam in the steam of total condensation or steam header in steam header 26, in containment 25, carry out condensation by containment cooling system again and form condensate water, condensate water falls into melt pit, and supplement chilled water to melt pit 105, maintain melt pit and flood liquid level absolute altitude, ensure the stability of the long-term cool cycles of reactor core.Heat in containment is discharged into by containment cooling system in the ambient atmosphere outside containment.
Fig. 1 is the PIPING OF MAIN LOOPS IN NUCLEAR POWER STATION system schematic of prior art.Fig. 5 is the schematic diagram of the non-active pressure release condensation heat exchange system according to another embodiment of the present utility model.As shown in Figure 1 and Figure 5, non-active nuclear power station comprises containment 25 and main automatic dropping valve 23, and main automatic dropping valve 23 is communicated with the major loop hot arc 4 be connected on reactor pressure vessel, for the steam in the release reaction core pressure vessel when having an accident.
Fig. 5 is the schematic diagram of the non-active pressure release condensation heat exchange system according to another embodiment of the present utility model.As shown in Figure 5, non-active nuclear power station pressure release condensation heat exchange system comprises the steam header 26 and closed natural convection loop that are communicated with main automatic dropping valve 23, closed natural convection loop comprises non-active vapor condensation heat-exchange device 27, heat-exchanging loop pipeline 31, the outer non-active heat exchanger 30 of shell and heat transfer medium, wherein steam header 26 is arranged in the containment 25 of non-active nuclear power station, the top of steam header is configured with steam header discharge of steam pipeline 108, the bottom of steam header is configured with the steam header condensed water elimination pipeline 110 being provided with retaining valve 29, heat-exchanging loop pipeline 31 runs through steam header 26 and containment 25 is arranged, non-active vapor condensation heat-exchange device 27 to be arranged in steam header 26 and to be communicated with heat-exchanging loop pipeline 31, it is outer and be communicated with heat-exchanging loop pipeline 31 that the outer non-active heat exchanger 30 of shell is arranged on containment 25.In this embodiment shown in Fig. 5, when the heat transfer medium in closed natural convection loop is the potpourri comprising cooling water and steam (can contain freeze-point depressant if desired), closed natural convection loop is evacuated (being such as 0.475 absolute atmosphere).
When main automatic dropping valve 23 is opened and the steam in reactor pressure vessel 2 is discharged into steam header 26, main automatic dropping valve 23 spurts the steam and is discharged in steam header 26 along the direction pointed by arrow F3, steam is through non-active vapor condensation heat-exchange device 27, to heat transfer medium (the such as chilled water in non-active vapor condensation heat-exchange device, can freeze-point depressant be contained if desired) carry out heating and forming condensate water by heat exchange, condensate water is discharged in melt pit 105 by the steam header condensed water elimination pipeline 110 and the retaining valve 29 be arranged on steam header condensed water elimination pipeline being configured in the bottom of steam header, thus supplement chilled water to melt pit 105, ensure the stability of the long-term cool cycles of reactor core.
Chilled water in closed natural convection loop when main automatic dropping valve 23 is opened and its temperature exceedes design temperature (being such as 80 degrees Celsius) of the closed Natural Circulation starting closed natural convection loop time added thermosetting steam at non-active vapor condensation heat-exchange device 27 place, along heat-exchanging loop pipeline 31 towards shell, outer non-active heat exchanger 30 flows steam in closed natural convection loop, heat is discharged into the atmosphere by the outer non-active heat exchanger 30 of shell, steam in closed natural convection loop forms condensate water after cooling and relies on gravity again to get back to non-active vapor condensation heat-exchange device 27 in the outer non-active heat exchanger 30 of shell, thus closed Natural Circulation is established in closed natural convection loop, to continue the reactor core waste heat of the reactor core remnants fission generation taken away in reactor pressure vessel when having an accident.
Can not drain in containment via the discharge of steam pipeline 108 at steam header top by the gas of incoagulability that contains of the steam in the steam of total condensation or steam header in steam header 26, in containment 25, condensation is carried out again by containment cooling system, condensate water falls into melt pit, and supplement chilled water to melt pit 105, maintain melt pit and flood liquid level absolute altitude, ensure the stability of the long-term cool cycles of reactor core.Heat in containment is discharged into by containment cooling system in the ambient atmosphere outside containment.
It is pointed out that in the embodiment such as shown in Fig. 5 and Fig. 4, in order to ensure discharge of steam and condensate water discharging timely, bleeder line 108 and water discharge line 110 can adopt the mode of many pipeline parallel connections.In addition, bleeder line 108 and water discharge line 110 also can adopt other possible embodiments, and this also will in protection domain of the present utility model.
It is to be noted, in the embodiment such as shown in Fig. 5 and Fig. 4, for guaranteeing that non-active nuclear power station pressure release condensation heat exchange system of the present utility model can normally work, non-active nuclear power station pressure release condensation heat exchange system of the present utility model can adopt the conventional design of antifreeze, fouling gas (such as air) exhaust etc., and they are not repeated herein as prior art.Relative to the mode of the Natural Circulation heat exchange of Fig. 4, the Natural Circulation heat exchange mode of Fig. 5 has heat conduction by gravity heat pipe heat exchanger principle and stablizes rapidly, and the advantage of requirement settled by the outer refrigeratory of shell without a special high position.In order to the thermal discharge efficiency of the outer non-active heat exchanger of stiffened shell, the hyperbolic-type air cooling tower form of existing mature technology can be adopted.
Have the following advantages according to non-active nuclear power station pressure release condensation heat exchange system of the present utility model: 1) can by outside reactor core Residual heat removal containment, the pressurized effect that in effective reduction containment, steam is assembled, decrease the load of containment cooling system, the chilled water gravity flow outside shell being conducive to realizing containment relies on cross-ventilated long-term cooling after draining simultaneously.2) adopt non-active vapor condensation heat-exchange device to carry out condensation to the steam that main automatic dropping valve gives off and become condensate water, condensate water returns melt pit by the condensed water elimination pipeline bottom steam tank, supplement chilled water to melt pit, guarantee that long-term melt pit recycle cooling is carried out sustainedly and stably.3) decrease discharge of steam in containment, reduce the back pressure of main automatic dropping valve discharge, be conducive to accelerating to reduce the pressure in pressure vessel, guarantee that main water supply tank drops into and sustained water injection in time, thus make reactor core be in safer flooding and the state of cooling.4) the non-active nuclear power station pressure release condensation heat exchange system of setting up and reactor core medium completely isolated, decrease the risk that radioactivity leaks outside.5) do not change existing non-active core cooling system structure, and non-active nuclear power station pressure release condensation heat exchange system of the present utility model adopts non-enabling fashion, rely on natural force to drive, keep original non-active design concept.
Claims (7)
1. a non-active nuclear power station pressure release condensation heat exchange system, wherein non-active nuclear power station comprises containment and main automatic dropping valve, main automatic dropping valve is communicated with the major loop hot arc be connected on reactor pressure vessel, for the steam in the release reaction core pressure vessel when having an accident, it is characterized in that, non-active nuclear power station pressure release condensation heat exchange system comprises the steam header and closed natural convection loop being arranged to be communicated with main automatic dropping valve, closed natural convection loop comprises non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, the outer non-active heat exchanger of shell and heat transferring medium, wherein steam header is arranged in the containment of non-active nuclear power station, the top of steam header is configured with steam header discharge of steam pipeline, the bottom of steam header is configured with the steam header condensed water elimination pipeline being provided with retaining valve, heat-exchanging loop pipeline runs through steam header and containment is arranged, non-active vapor condensation heat-exchange device to be arranged in steam header and to be communicated with heat-exchanging loop pipeline, it is outer and be communicated with heat-exchanging loop pipeline that the outer non-active heat exchanger of shell is arranged on containment, the outer non-active heat exchanger of shell is arranged on higher position relative to non-active vapor condensation heat-exchange device, heat transferring medium in closed natural convection loop is absorbed heat at non-active vapor condensation heat-exchange device and is transferred heat to the outer non-active heat exchanger of shell by heat-exchanging loop pipeline, thus at non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, closed Natural Circulation is established between the outer non-active heat exchanger of shell, to continue taking away reactor core remnants fission producing reactor core waste heat when non-active nuclear power station has an accident.
2. non-active nuclear power station pressure release condensation heat exchange system as claimed in claim 1, it is characterized in that, heat transfer medium in closed natural convection loop is the potpourri of cooling water and steam, closed natural convection loop is evacuated, when main automatic dropping valve is opened and when the steam in reactor pressure vessel is discharged into steam header, main automatic dropping valve spurts the steam and is discharged in steam header, main automatic dropping valve spurts the steam that through non-active vapor condensation heat-exchange device, heat transfer medium in non-active vapor condensation heat-exchange device is heated and forms condensate water by heat exchange, condensate water is discharged in melt pit by the steam header condensed water elimination pipeline and the retaining valve be arranged on steam header condensed water elimination pipeline being configured in the bottom of steam header, thus supplement chilled water to melt pit, ensure the stability of the long-term cool cycles of reactor core, chilled water in closed natural convection loop when main automatic dropping valve is opened and its temperature exceedes the design temperature of the closed Natural Circulation starting closed natural convection loop time added thermosetting steam at non-active vapor condensation heat-exchange device place, steam in closed natural convection loop is outer non-active heat exchanger flowing along heat-exchanging loop pipeline towards shell, heat is discharged into the atmosphere by the outer non-active heat exchanger of shell, steam in closed natural convection loop forms condensate water after cooling and relies on gravity again to get back to non-active vapor condensation heat-exchange device in the outer non-active heat exchanger of shell, thus closed Natural Circulation is established in closed natural convection loop, to continue the reactor core waste heat of the reactor core remnants fission generation taken away in reactor pressure vessel when having an accident.
3. non-active nuclear power station pressure release condensation heat exchange system as claimed in claim 1, it is characterized in that, closed natural convection loop also comprises the outer heat exchange isolation valve of shell, the outer heat exchange isolation valve of shell is arranged between non-active vapor condensation heat-exchange device and the outer non-active heat exchanger of shell along heat transferring medium flowing direction in closed natural convection loop, and the outer heat exchange isolation valve of shell is interlocked with main automatic dropping valve and opened; when main automatic dropping valve is opened and when the steam in reactor pressure vessel is discharged into steam header, steam is through non-active vapor condensation heat-exchange device, heat transferring medium in non-active vapor condensation heat-exchange device is heated and forms condensate water by heat exchange, condensate water is discharged in melt pit by the steam header condensed water elimination pipeline and the retaining valve be arranged on steam header condensed water elimination pipeline being configured in the bottom of steam header, flowed towards the outer non-active heat exchanger of shell along closed natural convection loop by the heat transferring medium heated in non-active vapor condensation heat-exchange device, heat is discharged into the atmosphere by the outer non-active heat exchanger of shell, in the outer non-active heat exchanger of shell, cooled heat transferring medium relies on gravity again to get back to non-active vapor condensation heat-exchange device, thus at non-active vapor condensation heat-exchange device, heat-exchanging loop pipeline, closed Natural Circulation is established between the outer non-active heat exchanger of shell and the outer heat exchange isolation valve of shell, to continue the reactor core waste heat of the reactor core remnants fission generation taken away in reactor pressure vessel when having an accident.
4. non-active nuclear power station pressure release condensation heat exchange system as claimed in claim 1, it is characterized in that, can not drain in containment via the discharge of steam pipeline at steam header top by the gas of incoagulability that contains of the steam in the steam of total condensation or steam header in steam header, the heat in containment be discharged into the atmosphere by containment cooling system.
5. non-active nuclear power station pressure release condensation heat exchange system as claimed in claim 1, it is characterized in that, the reactor core cooling system that non-active nuclear power station comprises primary heat transport system and is communicated with it, reactor core cooling system is used for taking away when having an accident the reactor core waste heat that in primary heat transport system, reactor core remnants fission produces.
6. non-active nuclear power station pressure release condensation heat exchange system as claimed in claim 5, it is characterized in that, primary heat transport system comprises steam generator, U-tube, cold section of major loop, major loop hot arc, main pump, reactor pressure vessel, be positioned at the reactor core of reactor pressure vessel, Surge line piping and voltage stabilizer, wherein U-tube is arranged in a vapor generator, U-tube endpiece is communicated with through cold section of main pump and major loop through the cold chamber compartment by steam generator bottom, cold section of major loop is communicated with reactor pressure vessel, reactor pressure vessel is communicated with major loop hot arc, major loop hot arc to be communicated with voltage stabilizer by Surge line piping and the hot chamber compartment passing through steam generator bottom is communicated with the inlet end of U-tube, cooling medium enters reactor pressure vessel by cold section of major loop, arrive the entrance of reactor core, the Q-value that reactor core produces is taken away when flowing through reactor core, major loop hot arc is flowed through by the cooling medium heated, arrive the hot chamber compartment of steam generator bottom and enter the inlet end of U-tube, to be transferred heat in steam generator by U-tube and cooling medium outside U-tube, coolant temperature in U-tube reduces and collects in the cold chamber compartment of steam generator bottom by the endpiece of U-tube, the main pump that cooling medium in cold chamber compartment is communicated with cold cavity bottom pumps into cold section of major loop, again get back to reactor pressure vessel, form the enclosed cool cycles of primary heat transport system.
7. non-active nuclear power station pressure release condensation heat exchange system as claimed in claim 5, it is characterized in that, reactor core cooling system comprises the first water supply tank, second water supply tank, main water supply tank, be arranged in the passive residual heat removal heat exchanger of main water supply tank, level Four Automatic Depressurization System, melt pit, melt pit filter screen, melt pit return line and the explosive valve be arranged on melt pit return line, first water supply tank, second water supply tank, main water supply tank is communicated with reactor pressure vessel by direct reactor peace note pipe respectively by corresponding connecting line and the retaining valve be arranged on each connecting line, first water supply tank top is communicated with by cold section of pressure-equalizing line and major loop, thus the pressure of the pressure in the first water supply tank and primary heat transport system is consistent, level Four Automatic Depressurization System comprises the automatic dropping valve of the first order, the automatic dropping valve in the second level, the automatic dropping valve of the third level and main automatic dropping valve, the automatic dropping valve of the first order, the automatic dropping valve in the second level, the inlet end of the automatic dropping valve of the third level to be connected on voltage stabilizer and the automatic dropping valve of the first order with parallel way, the endpiece of the automatic dropping valve of the second level automatic dropping valve third level is connected on main water supply tank with parallel way, main automatic dropping valve is communicated with major loop hot arc, passive residual heat removal heat exchanger is communicated with major loop hot arc with cold section of major loop, Natural Circulation is established at passive residual heat removal heat exchanger and between cold section of major loop and major loop hot arc, reactor pressure vessel is arranged in melt pit, chilled water in melt pit is by melt pit return line, melt pit filter screen be arranged on melt pit return line borehole blasting valve and be communicated with reactor pressure vessel by direct safety injection pipe.
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| CN105788676A (en) * | 2016-05-06 | 2016-07-20 | 上海核工程研究设计院 | Passive special safety facility of nuclear power station |
| CN105810256A (en) * | 2014-12-29 | 2016-07-27 | 国核华清（北京）核电技术研发中心有限公司 | Passive residual heat removal system for nuclear power plant |
| CN105810257A (en) * | 2014-12-29 | 2016-07-27 | 国核华清（北京）核电技术研发中心有限公司 | Pressure release condensation heat transfer system for passive nuclear power station |
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| CN109322718A (en) * | 2018-09-12 | 2019-02-12 | 西安交通大学 | The system and method for Nuclear Power System natural-circulation capacity is improved using residual heat of nuclear core |
| CN109322718B (en) * | 2018-09-12 | 2020-09-08 | 西安交通大学 | System and method for improving natural circulation capacity of nuclear power system by using reactor core waste heat |
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