Nuclear power plant

Abstract

To delete a risk of an excessive temperature damage of a reactor containment vessel, a risk of leakage of a large amount of hydrogen and a radioactive substance by deterioration of a seal material of a flange due to a high temperature of an upper lid part of the reactor containment vessel, and a risk of an early overpressure damage of the reactor containment vessel by a stratification of a temperature of pressure suppression pool water, and prevent the damage of the reactor containment vessel and the leakage of the radioactive substance without using a filter vent.SOLUTION: A nuclear power plant comprises a static containment vessel cooling system that has: a reactor containment vessel 3; a cooling water pool 13 that is provided in an upper part of the reactor containment vessel 3; a heat exchanger 16 of which a part is submerged into cooling water 14; condensed water return piping 21 that returns condensed water generated by the heat exchanger 16 to a pressure suppression pool 6; reactor pressure vessel water injection piping 51 that branches from the condensed water return piping 21 to inject the condensed water to an interior part of the reactor pressure vessel 2; gas vent piping 22 that ventilates gas generated within the heat exchanger 16 to the pressure suppression pool 6; and a gas vent fan 22a that is provided in the gas vent piping 22.SELECTED DRAWING: Figure 1

Description

[Technical field]

[0001] The present invention relates to nuclear power plants using passive containment cooling systems. [Background technology]

[0002] A conventional nuclear power plant and a conventional passive containment vessel cooling system will be outlined with reference to Figs.

[0003] (Figure 15: Explanation of conventional nuclear power plant No. 1) Fig. 15 is an elevational sectional view showing an example of the configuration around the reactor containment vessel of a first conventional nuclear power plant. In Fig. 15, a reactor core 1 is housed inside a reactor pressure vessel 2. The reactor core 1 is further housed inside a shroud 2a and a shroud head 2b (see Fig. 19). The reactor pressure vessel 2 is housed inside a reactor containment vessel 3. The reactor containment vessel 3 has a cylindrical shape (see Fig. 16). The inner diameter of the reactor containment vessel 3 is about 29 m. This configuration is adopted in the advanced BWR (ABWR).

[0004] The inside of the reactor containment vessel 3 is divided into a drywell 4, which contains the reactor pressure vessel 2, and a wetwell 5, and the drywell 4 and wetwell 5 together form part of the reactor containment vessel 3. The wetwell 5 forms a pressure suppression pool 6 inside. A wetwell gas phase 7 is formed above the pressure suppression pool 6. The outer walls of the drywell 4 and the wetwell 5 are integrated to form the cylindrical outer wall of the reactor containment vessel 3. The ceiling of the drywell 4 is a flat plate, and this part is called the top slab 4a of the drywell 4. The thickness of the top slab 4a is approximately 2.4 m.

[0005] In the case of a boiling water reactor, the atmosphere in the containment vessel 3 is replaced with nitrogen and the oxygen concentration is limited to a low level. In the case of a boiling water reactor, the containment vessel 3 is housed inside the reactor building 100.

[0006] Generally, the reactor containment vessel 3 is made of various materials, such as a steel reactor containment vessel, a reinforced concrete reactor containment vessel (RCCV), a prestressed concrete reactor containment vessel (PCCV), and a steel-concrete composite containment vessel (SCCV). In the case of RCCV and PCCV, a steel liner is lined on the inside. Figure 15 shows an example of an RCCV. As shown in Figure 16, the outer wall of the RCCV is cylindrical.

[0007] The reactor pressure vessel 2 is supported by a cylindrical pedestal 61 via an RPV skirt 62 and an RPV support 63. The pedestal 61 may be made of steel, concrete, or a composite structure of both. The space inside the pedestal 61, which is below the reactor pressure vessel 2 and the RPV skirt 62 in the drywell 4 and is surrounded by the cylindrical wall of the pedestal 61, is called the pedestal cavity 61a. In the case of the RCCV of the ABWR, the cylindrical wall of the pedestal 61 forms the boundary wall between the wetwell 5 and the drywell 4, and this space in particular is called the lower drywell 61a.

[0008] In the case of the RCCV of the ABWR, the space above the RPV support 62 of the drywell 4 is called the upper drywell 4c. The ceiling of the wetwell 5 is a flat plate and forms the boundary with the upper drywell 4c. This part forms the floor of the upper drywell 4c and is called the diaphragm floor 5b. The height from the upper end of the diaphragm floor 5b to the lower end of the top slab 4a is approximately 9m. The area below the diaphragm floor 5b of the reactor containment vessel 3 is called the lower region 90. The area above the diaphragm floor 5b and below the top slab 4a of the reactor containment vessel 3 is called the upper region 91. The lower region 90 and upper region 91 include the spaces, pools, and equipment provided in each region (see Figure 18).

[0009] A reactor containment vessel top head 10 is disposed above the reactor pressure vessel 2, and a water shield 11 is disposed above that.

[0010] A main steam pipe 71 extends from the reactor pressure vessel 2 to the outside of the dry well 4. A safety relief valve ("SRV") 72 is provided on the main steam pipe 71, and a relief pipe 73 is provided so as to be submerged in the suppression pool 6 so that steam from the reactor pressure vessel 2 is released into the suppression pool 6 when the safety relief valve 72 operates.

[0011] The dry well 4 and the suppression pool 6 are connected by a LOCA vent pipe 8. There are multiple LOCA vent pipes 8 installed, for example ten, but only two are shown in Figure 15 (see Figure 16). The LOCA vent pipe 8 has a horizontal vent pipe 8a in the part of the suppression pool 6 that is submerged in the pool water and opens into the pool water. In the case of RCCV, three horizontal vent pipes 8a are installed vertically on one LOCA vent pipe 8. In addition, in the case of RCCV, the LOCA vent pipe 8 is installed through the inside of the cylindrical wall of the pedestal 61. Therefore, in the case of RCCV, this cylindrical wall of the pedestal 61 is also called the vent wall. The vent wall is made of reinforced concrete with a thickness of about 1.7 m, and the inner and outer surfaces are made of steel. The LOCA vent pipe 8 and the pedestal 61 form part of the reactor containment vessel 3.

[0012] A vacuum breaker valve 9 is provided for the purpose of returning the gas in the wet well gas phase portion 7 to the dry well 4. A plurality of vacuum breaker valves 9, for example eight, are installed, but only one is shown in FIG.

[0013] The vacuum breaker valve 9 may be installed on the wall of the wetwell 5, on the ceiling of the wetwell 5, or on the LOCA vent pipe 8. The vacuum breaker valve 9 operates and opens when the pressure in the wetwell 5 becomes higher than the pressure in the drywell 4 and the differential pressure exceeds a set pressure. The set pressure of the vacuum breaker valve 9 is, for example, about 2 psi (about 13.79 kPa). The vacuum breaker valve 9 constitutes a part of the reactor containment vessel 3.

[0014] (Figure 17: Explanation of the configuration of the reactor containment vessel top head section) The configuration of the containment vessel top head part 10a will be described with reference to Fig. 17. The top head 10 of the containment vessel 3 is provided at the center of the top slab 4a of the dry well 4. A water shield 11 is provided on the top of the top head 10. Among already constructed plants, only Mark III containment vessels have this water shield 11, and concrete shielding blocks are provided in other BWR plants. Water shielding is used in the ESBWR, but has not yet been constructed at the time of filing the present invention.

[0015] The top lid 10 of the reactor containment vessel 3 is made of steel and has a structure that allows it to be removed from the flange 74 during fuel replacement. The flange 74 is circular along the opening of the top lid 10, but Figure 17 only shows cross sections of two places, one on the left and one on the right. A sealant (not shown) is inserted into the joint of the flange 74 to maintain airtightness. Tests have confirmed that the sealant has the ability to maintain airtightness for at least 168 hours at a high temperature of 200°C. The flange 74 is normally fastened with bolts (not shown) and the top lid 10 of the reactor containment vessel 3 is closed.

[0016] The upper part of the reactor pressure vessel 2, including the head 2c of the reactor pressure vessel 2, is housed inside the top lid 10 of the reactor containment vessel 3. A platform 75 is provided between the upper part of the reactor pressure vessel 2 and the reactor containment vessel 3. A bellows (not shown) is provided at the joint between the reactor containment vessel 2 and the platform 75, so that the joint can be flexibly maintained even if there is thermal expansion or displacement during an earthquake while maintaining watertightness. The space surrounded by the top lid 10 of the reactor containment vessel 3, the platform 75, and the part of the reactor pressure vessel 2 above the platform 75 is called the top lid 10a of the reactor containment vessel 3. There are several manholes 76 in the platform 75, which are normally open. Only one manhole is shown in FIG. 17. Since the opening area of ​​the manhole 76 is limited, the top lid 10a of the reactor containment vessel 3 is a space with a relatively high level of sealing.

[0017] Insulation material (not shown) is installed around the reactor pressure vessel 2 to suppress heat radiation during normal operation. Nevertheless, a certain amount of heat radiation occurs through the insulation material. During normal operation, heat is radiated from the head 2c of the reactor pressure vessel 2 to the top lid 10a. To prevent the temperature inside the top lid 10a from becoming too high during normal operation, several ventilation ducts 77 are provided penetrating the platform 75. Only one is shown in FIG. 17. Low-temperature gas (mainly nitrogen) is supplied to the ventilation duct 77 from a drywell cooler (not shown) to maintain the temperature inside the top lid 10a low. The heated gas is exhausted from the manhole 76 to the drywell 4 and returned to the drywell cooler. In this way, during normal operation, the gas inside the top lid 10a is forced to circulate and the temperature is maintained low. During fuel exchange, the manhole 76 and the ventilation duct 77 are closed to ensure watertightness, and the top lid 10 of the reactor containment vessel 3 can be opened to store water for shielding inside the top lid 10a.

[0018] (Figure 18: Explanation of the areas inside the reactor containment vessel 3) The area inside the containment vessel 3 will be described with reference to Fig. 18. The area inside the containment vessel 3 and below the diaphragm floor 5b is a lower area 90 of the containment vessel 3. The lower area 90 includes the wetwell 5, the wetwell gas phase 7, the pressure suppression pool 6, the LOCA vent pipe 8, the lower drywell 61a, and a core catcher (not shown).

[0019] The region inside the reactor containment vessel 3, which is above the diaphragm floor 5b and below the top slab 4a, is an upper region 91 of the reactor containment vessel 3. The upper region 91 includes a part of the upper dry well 4c, a part of the reactor pressure vessel 2, and a part of the equipment inside the reactor pressure vessel 2.

[0020] The space inside the reactor containment vessel 3 and surrounded by the top head 10, the platform 75, the head 2c of the reactor pressure vessel 2, and a part of the reactor pressure vessel 2 is the top head 10a of the reactor containment vessel 3. The top head 10a is a space and does not include the head 2 of the reactor containment vessel 2.

[0021] (Figure 19: Explanation of reactor pressure vessel 2 and reactor internals) The reactor pressure vessel 2 and reactor internals are explained based on Figure 19. Figure 19 shows an example of the reactor pressure vessel 2 and reactor internals of an advanced boiling water reactor (ABWR). The origin of the boiling water reactor (BWR) is the pressurized water reactor (PWR). The ABWR has achieved ultimate simplification and high output by removing the steam generator (SG), pressurizer, accumulator, hot leg piping, and cold leg piping from the PWR. Currently, a large PWR with four loops has four steam generators (SG), one pressurizer, four accumulators, four cold leg piping, and four hot leg piping, but these have all been removed in the ABWR. In the case of a large PWR, there are ten high pressure vessels for the primary system inside the reactor containment vessel, but in the ABWR, there is only one reactor pressure vessel 2 inside the reactor containment vessel 3. Moreover, the price of one SG is higher than the price of the reactor pressure vessel 2 of the ABWR because it contains the SG tubes, which are precision instruments. The plant cost of a large PWR is now at its limit, at approximately 1.6 trillion yen. This is the result of pushing for higher power output by simply increasing the number of loops without technological innovation. Therefore, in recent years, the development of small modular reactors (SMRs), which are simplified PWRs, has been actively carried out. When simplifying a PWR, if the external SG is removed and an internal SG is adopted, the internal SG must be made smaller because it is installed inside the reactor pressure vessel, and the reactor output must be reduced. Miniaturization is one of the problems with SMRs. Therefore, modularization has become necessary to increase power output. The output of an ABWR is approximately 1315 MWe net, but in order to obtain the same output as an ABWR with an SMR of 77 MWe net, approximately 17 modularization units are required. However, since the ABWR has removed the SG itself, there is no restriction on miniaturization, and it is possible to achieve ultimate simplification and high power output at the same time. Since there is no internal SG inside the reactor pressure vessel 2 of the ABWR, it is simpler than the reactor module of an SMR. Moreover, modularization is not required at all. A single reactor pressure vessel 2 can generate a net output of 1,315 MWe.

[0022] A shroud 2a and a shroud head 2b are provided inside the reactor pressure vessel 2. The shroud 2a and the shroud head 2b are integrated to form a watertight vessel. The core 1 is provided inside the shroud 2a and the shroud head 2b. The shroud 2a and the shroud head 2b divide the inside of the reactor pressure vessel 2 into two. The outer part of the side wall of the shroud 2a and the shroud head 2b of the reactor pressure vessel 2 is called the downcomer 2d. A feedwater sparger 50 is provided above the downcomer 2d. The feedwater sparger 50 has an annular shape, but for simplicity, only the cross section of one side is shown. Feedwater is supplied to the feedwater sparger from a feedwater piping (not shown). In this way, feedwater is supplied to the downcomer 2d and not directly to the core 1.

[0023] An internal pump 40 is provided under the shroud 2a. The internal pump 40 has a modular configuration with multiple identical pump modules. For example, in the case of a 1,350 MWe class ABWR, ten internal pumps 40 are provided. For simplicity, only two internal pumps 40 are shown in the figure. Incidentally, the ABWR can change its rated output by changing the number of internal pump 40 modules. Therefore, with an ABWR, it is no longer necessary to design the entire plant to have a modular configuration, such as by providing multiple modular reactor containment vessels 2 and 3.

[0024] Feedwater supplied from the feedwater sparger 50 to the downcomer 2d is supplied to the core 1 from the lower part of the shroud 2a by the internal pump 40. A pump deck 41 is provided between the lower part of the shroud 2a and the side wall of the reactor pressure vessel 2. A diffuser 42 for the internal pump 40 is provided penetrating the pump deck 41. An impeller (not shown) is provided inside the diffuser 42. The feedwater in the downcomer 2d is forcibly pushed into the shroud 2a by the impeller and diffuser 42.

[0025] A large number of steam-water separators 43 are provided on the upper part of the shroud head 2b. For simplicity, only two are shown in Figure 19. The interior of the steam-water separator 43 is configured so that a two-phase flow of water and steam flows in from the shroud head 2b. Furthermore, a steam dryer 44 and a steam dryer skirt 45 are provided above the steam-water separator 43. A large number of steam dryers 43 are installed, but for simplicity, only four are shown.

[0026] The water forced into the shroud 2a by the internal pump 40 rises and reaches the core 1, where a part of the water boils due to the heat generated by the nuclear fuel in the core 1, generating steam. The two-phase flow of water and steam rises inside the core 1 and the shroud head 2b and flows into the steam-water separator 43. The steam and water are separated by the steam-water separator 43, and the steam further flows into the steam dryer 44. The separated water flows out of the steam-water separator 43 and becomes reactor water. As a result, the reactor water level during normal operation is formed at approximately the middle position of the steam-water separator 43. The reactor water communicates with the inside and outside of the steam dryer skirt 45 and forms the same reactor water level. Steam exists in the space above the reactor water level, forming a steam phase 2s. The steam that flows into the steam dryer 44 has water droplets removed and becomes dry steam. This dry steam moves to the steam dome 2e and the steam phase 2s, and is supplied to the turbine (not shown) from the main steam pipe 71 (see FIG. 15). The steam pressure in the steam dome 2e also acts on the reactor water in the downcomer 2d. That is, the pressure of the steam phase 2s above the reactor water level in the reactor pressure vessel 2 is equalized by the steam pressure. Therefore, when the water in the downcomer 2d is injected into the reactor by the internal pump 40, the steam pressure in the reactor core 1 is offset by the steam pressure in the steam phase 2s and does not become a resistance pressure.

[0027] In this way, in an ABWR, only the steam generated in the reactor is supplied to the turbine, and most of the reactor water is separated in the steam separator 43, mixed with the feed water again, and recirculated inside the reactor pressure vessel 2. In addition, the flow rate of the coolant can be significantly reduced because the latent heat of vaporization caused by the phase change of the coolant is used to cool the core fuel. Therefore, the capacity of the reactor coolant pump can be significantly reduced compared to a PWR, in which all the reactor water is sent to the SG as water. As a result, the in-plant load factor of an ABWR is only 3%, which is significantly smaller than the approximately 7% to 10% of a PWR. In the case of a PWR, approximately 7% to 10% of the electricity is consumed for forced circulation of the coolant, etc.

[0028] A sparger 46 of the low pressure core injection system (LPFL) is located inside the containment vessel 2. The LPFL sparger 46 injects cooling water into the downcomer 2d. In addition, a sparger 47 of the high pressure core injection system (HPCF) is located inside the shroud head 2b. The spargers 46 and 47 are annular in shape, but for simplicity, only cross sections are shown.

[0029] (Figure 15: Explanation of the high pressure core injection system) The configuration of the high pressure core injection system (HPCF) 92 will be described with reference to FIG. 15. The high pressure core injection system 92 is configured to suck in water from the suppression pool 6 through a suction pipe 94, pressurize it with a high pressure core injection system pump 93, and spray water directly onto the core from the sparger 47 in the shroud head 2b through an injection pipe 95. This design of spraying cooling water from the high pressure core injection system 92 directly onto the core 1 in the event of a loss of coolant accident was made by the inventor of the present invention with the intention of stopping the boiling of the reactor water in the event of a loss of coolant accident. Hitachi and GE initially opposed this design, but after discussion, GE agreed to the design for a different reason and it was adopted. Hitachi opposed it until the end, but finally agreed. In addition to stopping boiling, the inventor was concerned about the removal of the core spray system from the ABWR on the grounds that the core would not be exposed in the event of a LOCA, and thought that the HPCF piping should be maintained inside the shroud in preparation for a situation in which a core spray system would be required. It was thought that even in the ABWR, a situation may occur in which it is necessary to spray cooling water directly onto the core fuel for reasons that are not yet clearly understood, that is, a situation in which the core is exposed. In addition, the ABWR has two high-pressure core injection systems 92, but for simplicity, only one system is shown in Figure 15. This is an increase from the one electrically driven high-pressure injection system up to the BWR / 6 to two systems. This was implemented with the intention of reducing the operation frequency of the automatic depressurization system (ADS) (not shown) and reducing the occurrence frequency of a high-pressure scenario in the probabilistic safety assessment (PRA) called TQUX, thereby reducing the core meltdown frequency (CDF) of the ABWR. The effect of reducing the operation frequency of the ADS has a safety value that cannot be recognized in a PRA that only evaluates the CDF value. It is effective in reducing the stress of the operator during transient changes and preventing the occurrence of human error. This was also implemented by the inventor of the present invention. An outer isolation valve 96 and an inner isolation valve 97 are provided at the penetration part of the dry well 4 of the injection pipe 95. The outer isolation valve 96 is a motorized valve and the inner isolation valve 97 is a check valve.

[0030] (Figure 15: Explanation of the PCCS configuration. For the conventional first nuclear power plant) The configuration of a passive containment cooling system (PCCS) 12 in a conventional first nuclear power plant will be described with reference to Figures 15 and 20. A cooling water pool 13 is provided at the top of the reactor containment vessel 3, and cooling water 14 is stored inside. The cooling water pool 13 can be of a pool type or a tank type. Figure 15 shows a tank type. In the case of the pool type, the top is covered with a top lid (see Figure 21). Figure 15 shows an example in which the cooling water pool 13 etc. is installed inside the reactor building 100, but it may also be installed inside an adjacent auxiliary building etc.

[0031] An exhaust port 15 is provided to release steam from the gas phase above the water surface of the cooling water pool 13 into the environment. An insect screen may be provided at the outlet of the exhaust port 15. The cooling water pool 13 is generally located above the reactor containment vessel 3, but can also be located next to the reactor containment vessel 3.

[0032] A heat exchanger 16 is installed in the cooling water pool 13 so that at least a part of it is submerged in the cooling water 14. In many cases, multiple heat exchangers 16 are installed, but only one is shown in Figure 15. The heat exchanger 16 has an inlet plenum 17, an outlet plenum 18, and heat transfer tubes 19 (see Figure 20). The heat exchangers 16 as a whole have a cooling capacity to condense steam at least equivalent to the decay heat.

[0033] 15 shows an example in which only the heat transfer tubes 19 are installed inside the cooling water pool 13, and the inlet plenum 17 and the outlet plenum 18 are installed protruding outside the cooling water pool 13, but the example is not limited to this. For example, there is also an example in which the entire heat exchanger 16, including the inlet plenum 17 and the outlet plenum 18, is installed inside the cooling water pool 13 (see FIG. 21).

[0034] A drywell gas supply pipe 20 that supplies gas within the drywell 4 is connected to the inlet plenum 17 of the heat exchanger 16. One end of the drywell gas supply pipe 20 is connected to the drywell 4.

[0035] In addition, a condensate return pipe 21 is connected to the lower part of the outlet plenum 18 of the heat exchanger 16 (see FIG. 20). One end of the condensate return pipe 21 is led to the inside of the LOCA vent pipe 8, as an example, but is not limited to this example. There is also an example where the condensate return pipe 21 is led to a core catcher (not shown) at the lower part of the lower drywell 61a. Even when led to the inside of the LOCA vent pipe 8, the condensate is finally returned from the molten valve 64 to the core catcher (not shown) at the lower part of the lower drywell 61a.

[0036] In this way, the condensed water in the condensed water return pipe 21 is intended to be used to cool the core debris that falls onto the floor of the lower D / W 61a or onto the core catcher (not shown) in the event of a severe accident. This was based on the mistaken idea of ​​core catcher supremacists that in the event of a severe accident, it is most important to cool the debris that falls onto the floor of the lower dry well 61a or onto the core catcher (not shown) and prevent the concrete mat of the lower dry well 61a from melting and penetrating, and that this alone is sufficient (see Patent Documents 2 and 3). However, the lesson learned from the Fukushima Daiichi accident made it clear that in the event of a severe accident, the floor of the pedestal cavity 61a will not melt and penetrate, but rather it is important to take measures against the overheating and damage of the reactor containment vessel 3 or the deterioration of the sealing material of the flange 74 on the top lid 11 of the reactor containment vessel 3, causing a large amount of hydrogen and radioactive materials to leak.

[0037] In addition, a gas vent pipe 22 is connected to the upper part of the outlet plenum 18 (see FIG. 20 ) of the heat exchanger 16. One end of the gas vent piping 22 is led into the wet well 5 and installed so as to be submerged in the suppression pool 6. The gas vent piping 22 is installed so that the submersion depth in the suppression pool 6 of the gas vent piping 22 is shallower than the submersion depth of the uppermost end of the opening of the LOCA vent pipe 8 in the suppression pool 6.

[0038] (Figure 15: Explanation of flooded piping) In this conventional example, in preparation for a core meltdown during a transient event such as a station blackout (SBO), a meltdown valve 64 and a lower drywell flooding pipe 65 are provided inside the lower drywell 61a, which penetrate the wall of the pedestal 61 from the LOCA vent pipe 8 and connect to the meltdown valve 64. The meltdown valve 64 and the lower drywell flooding pipe 65 are provided on all of the LOCA vent pipes 8 (see FIG. 16). When the temperature of the lower drywell 61a reaches approximately 260°C, the meltdown valve 64 melts and opens. In the event of a core meltdown accident, the molten core melts through the bottom of the reactor pressure vessel 2 and falls into the pedestal cavity 61a. When the temperature in the pedestal cavity 61a rises sharply, the meltdown valve 64 opens, and the cooling water in the LOCA vent pipe 8 flows into the pedestal cavity 61a through the lower drywell flooding pipe 65 to flood and cool the molten core. This melting valve 64 design is used in the US-ABWR.

[0039] (Figure 20: Explanation of PCCS heat exchanger) Fig. 20 is an elevational sectional view showing an example of a heat exchanger in the passive containment cooling system in the conventional first nuclear power plant. The structure of the heat exchanger 16 in the passive containment cooling system 12 in the conventional first nuclear power plant will be described using an example of a horizontal heat exchanger with reference to Fig. 20.

[0040] In FIG. 20, the outlet plenum 18 is provided below the inlet plenum 17. A large number of U-shaped heat transfer tubes 19 are connected to a tube sheet 23, and the straight tube portions of the heat transfer tubes 19 are installed horizontally. For simplification, FIG. 20 shows only two tubes. The outside of the heat transfer tubes 19 is filled with cooling water 14 (see FIG. 15). The inlets of the heat transfer tubes 19 open to the inlet plenum 17. The outlets of the heat transfer tubes 19 open to the outlet plenum 18.

[0041] A drywell gas supply pipe 20 is connected to the inlet plenum 17, and supplies a mixed gas of nitrogen, hydrogen, water vapor, etc. in the drywell 4 to the inlet plenum 17. This mixed gas is led into the heat transfer tube 19, and the water vapor is condensed to become condensed water, which flows out from the outlet of the heat transfer tube 19 into the outlet plenum 18 and accumulates at the bottom of the outlet plenum 18.

[0042] A condensed water return pipe 21 is connected to the bottom of the outlet plenum 18, and the condensed water in the outlet plenum 18 is returned by gravity to the LOCA vent pipe 8 in the reactor containment vessel 3. In addition, a gas vent pipe 22 is connected to the top of the outlet plenum 18. Non-condensable gases such as nitrogen, hydrogen, and oxygen that do not condense in the heat transfer tubes 19 are discharged from the heat transfer tubes 19 and accumulate in the top of the outlet plenum 18.

[0043] The end of the gas vent piping 22 is led to the suppression pool 6, and the non-condensable gas in the outlet plenum 18 passes through the gas vent piping 22, pushes down the pool water in the suppression pool 6, and is vented into the pool water, after which it moves to the wet well gas phase section 7.

[0044] The shape of the heat transfer tubes 19 is not limited to a U-shape. There are also structures in which heat transfer tubes 19 with a straight tube section in the vertical direction are installed vertically (see Figure 25). The inlet plenum 17 is always located above the outlet plenum 18. This allows the condensed water that condenses inside the heat transfer tubes 19 to be guided to the outlet plenum 18 by gravity. The advantages of the horizontal type are excellent earthquake resistance and the ability to make effective use of the cooling water 14. On the other hand, the advantage of the vertical type is that the condensed water can be easily drained.

[0045] (Description of PCCS functions) Next, the function of the conventional passive containment cooling system 12 configured as above during a severe accident will be described.

[0046] When a severe accident occurs and damage to the core 1 begins inside the reactor pressure vessel 2, a metal-water reaction occurs between the hot core fuel cladding tubes and the cooling water, generating a large amount of hydrogen. This hydrogen is released to the pressure suppression pool 6 by the safety relief valve 72 and moves to the wetwell gas phase 7. The wetwell gas phase 7 is replaced with nitrogen to keep the oxygen concentration low, so hydrogen detonation does not occur. Hydrogen is added to the nitrogen that was present from the beginning, and the pressure in the wetwell gas phase 7 increases.

[0047] Furthermore, as the melting of the core fuel progresses, the bottom of the reactor pressure vessel 2 melts and debris (remnants of the molten core 1 that have become hot due to decay heat) falls to the bottom of the pedestal cavity (lower dry well) 61a. The heat of the debris opens the melting valve 64, and the water in the LOCA vent pipe 8 flows into the lower dry well 61a through the lower dry well flooding pipe 65, flooding the debris. The large amount of steam generated at this time flows into the upper dry well 4c from the opening 66 of the LOCA vent pipe 8 due to its own pressure, and then passes through the dry well gas supply pipe 20 and is led to the heat exchanger 16 of the passive containment cooling system 12 where it condenses. At this time, the non-condensable gas remaining in the heat exchanger 16 is vented into the wet well 5 through the gas vent pipe 22. In this state, the pressure in the dry well 4 is higher than the pressure in the wet well 5, so the non-condensable gas is efficiently vented to the wet well gas phase 7. The condensed water is returned to the LOCA vent pipe 8 of the reactor containment vessel 3 through the condensed water return pipe 21, passes through the melting valve 64, and is used again to cool the debris that has fallen to the bottom of the lower dry well 61a.

[0048] The debris that fell to the bottom of the lower dry well 61a has decay heat, which heats the water that has accumulated at the bottom of the lower dry well 61a and continues to generate steam. This steam continues to flow into the heat exchanger 16 of the passive containment cooling system 12 through the dry well gas supply pipe 20 under its own pressure, and is cooled by the heat transfer tube 19. This was supposed to cool the debris that fell to the lower D / W and cool the reactor containment vessel 3.

[0049] (Explanation of heating caused by remaining debris in the reactor) However, as a result of the Fukushima Daiichi accident, it has become clear that not 100% of the debris in the reactor falls into the lower D / W, but about 20% of the debris remains in the reactor. This means that the conventional passive containment cooling system 12 can only cool about 80% of the debris that fell into the lower dry well 61a. On the other hand, about 20% of the debris remaining in the reactor is not submerged and continues to emit decay heat while in a dry-fire state, causing it to become hot. As a result, the temperature inside the reactor pressure vessel 2 rises to about 1000°C. As the reactor pressure vessel 2 becomes hot and continues to emit heat, the temperature inside the dry well 4 may exceed 200°C for a long period of time. In other words, it has become clear that there are cases in which the reactor containment vessel 3 is not sufficiently cooled.

[0050] In particular, the top lid 10a of the reactor containment vessel 3 is partitioned by the platform 75, making it a narrow and poorly ventilated area. During normal operation, low-temperature gas is blown from the ventilation duct 77, so the temperature of the top lid 10a is kept low. However, in the event of a severe accident due to a total power outage (SBO), the drywell cooler (not shown) stops due to the loss of power, and the temperature inside the top lid 10a may exceed 300°C for a long period of time due to heat radiation from the head 2c of the reactor pressure vessel 2. The PCV top head 10 has a flange 74. A sealant (not shown) is inserted into the joint of the flange 74 to maintain airtightness. If the sealant of the flange 74 exceeds 200°C, it may deteriorate and cause a large amount of gas, such as hydrogen, inside the reactor containment vessel 3 to leak. The PCV top head 10 is cooled by the water shield 11, so the temperature of the outer surface does not greatly exceed 100°C. However, it is not preferable for the internal temperature of the PCV top head 10a to continue to exceed 300°C.

[0051] In the case of the Fukushima Daiichi accident, there was no water shielding 11 and the PCV top head 10 was covered with a concrete block shield, so it is believed that the PCV top head part 10a became abnormally hot, causing the sealing material of the flange 74 to deteriorate, resulting in a large amount of hydrogen gas, radioactive materials, etc. leaking onto the operating floor.

[0052] After that, as a voluntary measure, not required by the Nuclear Regulation Authority (NRA), it was considered to spray water from outside onto the top of the PCV head 10 using portable equipment, but it is possible that cooling from the outside alone is not enough. The gas temperature of the PCV head 10a must also be lowered. In order to lower the temperature of the PCV head 10a, it is necessary to cool the debris remaining in the reactor pressure vessel 2, which is the heat dissipation source. It is unclear whether this is the reason, but the NRA requires the provision of a cooling function for the molten core in the reactor pressure vessel 2 (probably a low-pressure water injection system is sufficient for PWR rescue) in its interpretation of Article 42 of the New Regulatory Requirements. This is because an alternative in-reactor high-pressure water injection system should have been installed as an additional measure to prevent core meltdown in the event of an SBO. This was possible in BWRs, where the reactor pressure was low, but the additional installation of such an alternative high-pressure water injection system was difficult in PWRs, where the reactor pressure was high, so it is thought that this was a compromised PWR rescue measure that relaxed the regulatory requirements and compromised by the additional installation of an alternative low-pressure water injection system after core meltdown. However, this remedy would be extremely effective as a means of cooling the debris remaining inside the reactor pressure vessel 2. It would be wonderful if the NRA had realized this.

[0053] In addition, plants that do not use core catchers and instead use In-vessel Retention (IVR) (a method in which the bottom of the reactor pressure vessel is cooled from the outside to prevent melting and damage of the reactor pressure vessel, but the debris inside the reactor pressure vessel is not directly cooled and is kept in a dry-fire state) do not have the cooling function for the molten core inside the reactor pressure vessel after the core melts, so they do not meet the new regulatory standards of Japan's NRA. This not only means that the NRA's regulatory requirements are not met, but also indicates that although IVR can prevent melting and penetration of the reactor pressure vessel in the event of a severe accident, there is a risk of the reactor containment vessel overheating and being damaged by heat from the molten core that is dry-fired inside the reactor pressure vessel. In the case of IVR, heat is generated not from about 20% of the debris remaining in the reactor, but from 100% of the debris remaining in the reactor, which continues to generate heat, causing a local rise in the temperature inside the reactor containment vessel, deteriorating the sealing material of the equipment hatch and causing a large amount of hydrogen and radioactive materials to be released to the outside. This is one of the important lessons learned from the Fukushima Daiichi accident. The lessons learned from the Fukushima Daiichi accident show that measures against overheating are more important than just melting and penetration. The lessons learned show that IVR can cause overheating of the containment vessel. However, an analytical model that can correctly handle the presence of debris remaining in the PWR core in a physical phenomenon analysis code during SA has not yet been developed. Therefore, even at the time of filing this patent application, there is a risk that the risk of overheating of the containment vessel due to debris remaining in the core may be overlooked in the development of innovative PWR reactors.

[0054] Unlike the AP1000 and EPR, existing Japanese PWRs do not have a means to depressurize the reactor pressure vessel in the event of an SBO. Therefore, even if alternative low-pressure water injection equipment is installed, water cannot be injected into the reactor until a core meltdown occurs, the bottom of the reactor pressure vessel is damaged, and the reactor pressure is reduced. Therefore, it appears that the NRA's requirement is not to prevent a core meltdown, but to provide a cooling function for the molten core inside the reactor.

[0055] However, in the case of BWRs, it is possible to install alternative high-pressure water injection equipment, and in the event of an SBO, the reactor pressure vessel 2 can be depressurized by the automatic depressurization system (ADS) and water can be injected using alternative low-pressure water injection equipment. Therefore, if these alternative water injection equipment are installed, core meltdown can be prevented. However, since the NRA's requirements are aimed at rescuing PWRs, the function is to cool the molten core inside the reactor after a core meltdown. Therefore, in the case of BWRs, even if alternative water injection equipment is installed, it is possible to prevent core meltdown with it, but these alternative water injection equipment can only be used after a core meltdown has occurred. Therefore, in BWRs, these alternative water injection equipment must be set to start only after a core meltdown has occurred and the radiation level inside the reactor containment vessel 3 has become abnormally high, which is a foolish thing to do. In order to match PWRs, which cannot prevent core meltdown even with alternative water injection equipment, BWRs, which can prevent core meltdown with alternative water injection equipment, interlock the alternative water injection equipment so that it cannot be started until a core meltdown occurs, deliberately causing a core meltdown. What a foolish outcome! However, this is not actually the fault of the NRA or PWR. The fault lies in the fixed idea of ​​using an alternative dynamic water injection system to cool the remaining debris inside the reactor pressure vessel 2. The NRA requires a cooling function for the molten core inside the reactor, but does not necessarily require an alternative dynamic water injection system. It requires some method to provide a cooling function for the molten core inside the reactor. This invention teaches an ideal solution to this requirement.

[0056] (Explanation of stratification of suppression pool water temperature) Another phenomenon that became clear from the results of the Fukushima Daiichi accident is the stratification of the suppression pool water temperature. The following will explain the case of a plant that did not have the passive containment cooling system 12 installed at the time of the Fukushima Daiichi accident. When a core meltdown accident occurs, a large amount of hydrogen is released from the relief valve 72 into the suppression pool 6 due to the metal-water reaction of the core fuel, and moves to the wetwell gas phase, causing the pressure in the wetwell gas phase to rise. Furthermore, when debris falls to the bottom of the reactor pressure vessel 2, the bottom of the reactor pressure vessel 2 is damaged and the debris falls to the floor of the lower drywell 61a. The heat of the debris activates the meltdown valve 64, and water flows into the lower drywell 61a from the LOCA vent pipe 8. Then, a large amount of high-pressure steam is generated from the water due to the heat of the debris. The steam moves from the opening 66 to the upper drywell 4c, pushes down the water level in the LOCA vent pipe 8, and moves from the horizontal vent pipe 8a into the suppression pool 6, where it is cooled by the pool water and condenses. At this time, the nitrogen present in the drywell 4 and the remaining hydrogen generated by the metal-water reaction of the core fuel are vented together with the steam into the pressure suppression pool 6 and move to the wetwell gas phase 7, causing the pressure in the wetwell gas phase 7 to rise further to near its limit.

[0057] Through this process, the decay heat of the debris that has fallen to the floor of the lower dry well 61a is transferred to the water in the suppression pool 6, causing the water to heat up. During this process, it has become clear that because the water in the suppression pool 6 is deep at approximately 7m, the hotter water moves to the top while the cold and hot water sinks to the bottom, causing stratification of the water temperature from the bottom to the top. In fact, stratification of the suppression pool water temperature was confirmed during the Fukushima Daiichi accident.

[0058] When the suppression pool water temperature stratification occurs, the pool water on the water surface becomes saturated in a short period of time and generates high-pressure steam. As a result, high-pressure saturated steam is generated in the wetwell gas phase 7 even though the average temperature of the entire pool water has not yet reached saturated water. A large amount of hydrogen and nitrogen in the drywell 4 have already migrated to the wetwell gas phase 7, and the pressure of these non-condensable gases has increased the pressure to near its limit. The high-pressure saturated steam generated from the pool water on the water surface of the suppression pool 6 causes the pressure in the wetwell gas phase 7 to become abnormally high. Furthermore, in conjunction with the pressure rise in the wetwell gas phase 7, the pressure in the drywell 4 also becomes high and exceeds the limit value (the pressure in the drywell 4 is higher than the pressure in the wetwell 5 by the head of water that vents the horizontal vent pipe 8a). In this state, there is a risk of overpressure damage to the reactor containment vessel 3, so the NRA requires the installation of a filter vent system. The filter vent system is a dangerous facility that directly releases radioactive organic iodine and radioactive rare gases generated in the reactor containment vessel 3 into the environment. To avoid this danger, local residents need to evacuate before the filter vent system is activated. However, it has become clear that due to the stratification of the suppression pool water temperature, the timing at which the filter vent system must be activated may be abnormally early, and the filter vent system may be activated or the reactor containment vessel 3 may be overpressurized and damaged before local residents have been able to evacuate completely. The present invention teaches a good solution to this problem.

[0059] (Explanation of temperature rise due to FP heating) For example, assuming that a main steam line rupture accident (MSLBA) occurs in which the main steam line (MSL) 71 ruptures in the drywell 4 and then progresses to a core meltdown accident, a large amount of radioactive fission products generated from the molten core will be released into the drywell 4 from the rupture in the main steam line. Most of the particulate radioactive material will plate out or fall out and deposit on the walls and floor of the drywell 4. Gaseous radioactive material and airborne radioactive material will be vented together with steam from the LOCA vent pipe 8 to the suppression pool 6. However, it has been found that the decay heat of a large amount of fission products adhering to the walls and floor of the drywell 4 will cause the ambient temperature of the drywell 4 to rise to about 400°C. The drywell 4 has an equipment hatch (not shown), and if the ambient temperature rises for a long period of time, the sealing material will deteriorate, causing a large amount of hydrogen and radioactive material in the drywell 4 to leak out. The present invention teaches a good solution to this problem.

[0060] (Figure 21: Explanation of a conventional second nuclear power plant) A second conventional nuclear power plant will be described with reference to Figs. FIG. 21 is an elevational sectional view showing an example of the configuration around the reactor containment vessel of the conventional second nuclear plant. This conventional example is shown in FIG. 9 and FIG. 2 of Patent Document 1. In FIG. 21, a reactor core 1 is housed inside a reactor pressure vessel 2. The reactor pressure vessel 2 is housed inside a reactor containment vessel 3. The reactor containment vessel 3 has a cylindrical shape (see FIG. 22). In FIG. 21, the configurations of the reactor core 1, the reactor pressure vessel 2, and the reactor containment vessel 3 are the same as those of the conventional first nuclear plant shown in FIG. 15. In the following, to avoid duplication, only the configurations in which the conventional second nuclear plant is different from the conventional first nuclear plant will be described.

[0061] (External well description) In the second conventional nuclear power plant, an outer well 32 is further provided outside the dry well 4 and the wet well 5. This configuration is adopted in the Severe Accident Tolerant Reactor (SATOR). The outer well 32 is adjacent to the dry well 4 via the dry well common wall 4b, and is adjacent to the wet well 5 via the wet well common wall 5a. The ceiling of the outer well 32 is a flat plate, and this portion is called the top slab 32a of the outer well 32. The atmosphere within the outer well 32 is replaced with nitrogen, and the oxygen concentration is limited to a low level. The outer well 32 has the same pressure resistance and airtightness as the dry well 4 and the wet well 5.

[0062] The material of the external well 32 can be any of the same materials as the reactor containment vessel 3, such as reinforced concrete (RC), SC, or steel. If it is made of reinforced concrete, a liner is laid on the inner surface, as in the case of the reactor containment vessel 3. As shown in FIG. 22, the planar shape of the external well 32 in this conventional example is rectangular and configured to surround part of the outer walls of the dry well 4 and the wet well 5, but the planar shape of the external well 32 is not limited to this. Any shape may be used as long as it is adjacent to and surrounds at least part of the outer walls of the dry well 4 and the wet well 5. For example, it may be a trapezoid, a polygon, a crescent shape, a partial ring shape, a complete ring shape, etc.

[0063] (Scrubbing pool explanation) The configuration of the scrubbing pool will be described with reference to Figure 23. A scrubbing pool 33 containing water is provided within an external well 32, and the top is covered with a top lid 33a, leaving a space 33b between the top lid 33a and the pool water. A first outlet pipe 33c opening into the space 33b is provided above the top lid 33a. A metal fiber filter 34 is connected to one end of the first outlet pipe 33c. In addition, there is a second outlet pipe 34a connected to the metal fiber filter 34 and the other end opening into the interior of the external well 32.

[0064] The space 33b is required in the event that the water level rises when gas in the reactor containment vessel 3 is vented from the gas vent pipe 22. The top cover 33a is also required to prevent water from flowing out due to sloshing during an earthquake. It is necessary to prevent water from flowing out due to sloshing when an earthquake occurs not only during normal times but also during accidents. In this conventional example, the shape of the scrubbing pool 33 is rectangular, but is not limited to this example. For example, any shape such as a polygon, ellipse, or circle may be used. There is also an example in which the scrubbing pool 33, the top cover 33a, and the space 33b are integrated into a tank.

[0065] (Figure 21: Explanation of the PCCS configuration. In the case of a conventional second nuclear power plant) Next, the configuration of the passive containment cooling system (PCCS) 12 of the conventional second nuclear power plant will be described with reference to Figures 21 and 24. A cooling water pool 13 is provided above the reactor containment vessel 3 and the external well 32, and stores cooling water 14 inside. The cooling water pool 13 is of a pool type and a tank type. Figure 21 shows a pool type. In the case of the pool type, the top is covered with a top lid.

[0066] An exhaust port 15 is provided to release steam from the gas phase above the water surface of the cooling water pool 13 into the environment. An insect screen may be provided at the outlet of the exhaust port 15. The cooling water pool 13 is generally located above the reactor containment vessel 3 and the external well 32, but may also be located next to the reactor containment vessel 3 and the external well 32.

[0067] A heat exchanger 16 is installed in the cooling water pool 13. The heat exchanger 16 is installed so that at least a part of it is submerged in the cooling water 14. In this conventional example, the heat exchanger 16 is installed so that it is completely submerged in the cooling water 14.

[0068] Although multiple heat exchangers 16 are often installed, only one is shown in Fig. 21. The heat exchanger 16 has an inlet plenum 17, an outlet plenum 18, and heat transfer tubes 19 (see Fig. 24). The heat exchangers 16 as a whole have a cooling capacity to condense steam at least equivalent to the decay heat.

[0069] 21 shows an example in which the entire heat exchanger 16, including the inlet plenum 17 and the outlet plenum 18, is installed inside the cooling water pool 13, but the present invention is not limited to this example. For example, there is also an example in which only the heat transfer tubes 19 are installed inside the cooling water pool 13, and the inlet plenum 17 and the outlet plenum 18 are installed protruding outside the cooling water pool 13.

[0070] A wetwell gas supply pipe 48 is connected to the inlet plenum 17 of the heat exchanger 16. One end of the wetwell gas supply pipe 48 opens into the wetwell gas phase part 7, penetrates the side wall of the wetwell 5, passes through the inside of the outer well 32, penetrates the top slab 32a of the outer well 32, and is connected to the inlet plenum 17 at the other end. In this conventional example, a drywell gas supply pipe 20 is also provided. One end of the drywell gas supply pipe 20 opens into the inside of the drywell 4, penetrates the side wall of the drywell 4, passes through the inside of the outer well 32, penetrates the top slab 32a of the outer well 32, and is connected to the wetwell gas supply pipe 48 at the other end. An isolation valve 20a is provided in the drywell gas supply pipe 20, and is closed during normal operation.

[0071] In addition, a condensed water return pipe 21 is connected to the lower part of the outlet plenum 18 of the heat exchanger 16. The condensed water return pipe 21 is installed so that it penetrates the top slab 32a of the external well 32, passes through the inside of the external well 32, and has its tip immersed in the pressure suppression pool 6 inside the wet well 5. Since the condensed water return pipe 21 is installed so as to pass through the inside of the external well 32 in this manner, the structure is such that condensed water does not leak and radioactive materials such as CsI are not directly released into the environment.

[0072] In the event of a core meltdown accident, the atmosphere within the reactor containment vessel 3 will contain large amounts of particulate radioactive material such as CsI, but most of the particulate radioactive material such as CsI will transfer to the condensed water when the steam condenses in the heat exchanger 16. The condensed water containing a large amount of CsI will be returned to the pool water of the suppression pool 6 by the condensed water return pipe 21 and will be retained therein, so the passive containment vessel cooling 12 of this conventional example is configured to statically remove particulate radioactive material suspended within the reactor containment vessel 3.

[0073] Therefore, in this conventional example, even if a loss of all AC power (SBO) leads to a core meltdown accident, it is possible to obtain an effect equivalent to that of removing particulate radioactive materials by dynamic containment vessel spray and returning them to the pool water in the suppression pool 6. In addition, since the condensate return pipe 21 is not installed inside the LOCA vent pipe 8, the structure does not increase the pressure loss of the LOCA vent pipe in the event of a LOCA.

[0074] Furthermore, a gas vent pipe 22 is connected to the upper part of the outlet plenum 18 of the heat exchanger 16. The gas vent pipe 22 passes through the inside of the external well 32 by penetrating the top slab 32a of the external well 32, and the end of the gas vent pipe 22 is further immersed in the water of the scrubbing pool 33. Since the gas vent pipe 22 is installed so as to pass through the inside of the external well 32, the structure is such that radioactive rare gases, organic iodine, CsI, and other radioactive substances are not directly released into the environment due to gas leakage. Among these, particulate radioactive substances such as CsI are removed by the pool water of the scrubbing pool 33, and further, any carried over in water droplets or the like is removed by the metal fiber filter 34.

[0075] This makes it possible to eliminate the need for long-term relocation of surrounding residents due to land contamination caused by particulate radioactive materials such as CsI being released into the environment. Also, radioactive rare gases and organic iodines are released from the second outlet pipe 34a into the inside of the external well 32 and are retained inside the external well 32. This makes it possible to eliminate the need for surrounding residents to evacuate in advance or take iodine tablets, as conventional filter-vent systems release radioactive rare gases and organic iodines directly into the environment when activated.

[0076] In addition, in this conventional example, hydrogen generated in large quantities by metal-water reactions in the fuel cladding tubes during a severe accident is released into the outer well 32 through the gas vent piping 22, making it possible to keep the pressure in the dry well 4 and the wet well 5 sufficiently low during a severe accident. Since the atmosphere inside the outer well 32 is replaced with nitrogen, there will be no explosion even if a large amount of hydrogen is trapped.

[0077] In addition, in this conventional example, since steam generated in the wetwell gas phase 7 can be sucked in through the wetwell gas supply piping 48 and cooled by the passive containment cooling system 12, even if the water temperature in the suppression pool 6 becomes stratified during a severe accident and high-pressure steam is generated from the water at the surface, the pressure in the reactor containment vessel 3 can be kept low and overpressure damage can be prevented. This is because the non-condensable gas in the reactor containment vessel 3 can be sucked in through the wetwell gas supply piping 48 and vented through the gas vent piping 22 into the scrubbing pool 33 and further into the outer well 32. In the case of the passive containment cooling system 12 of the conventional first nuclear power plant (ABWR) in which the gas vent pipe 22 is led into the suppression pool 6 (see Figures 15 and 20), even if a wetwell gas supply pipe 48 is provided, the gas cannot be vented to the wetwell 5 by the gas vent pipe 22, and non-condensable gas in the wetwell gas phase 7 accumulates inside the heat exchanger 16 of the passive containment cooling system 12, causing the passive containment cooling system 12 to lose function. In the case of the passive containment cooling system 12 of the conventional second nuclear power plant (SATOR) (see Figures 21 and 24), the outer well 32 is provided as the vent destination of the gas vent pipe 22, so that non-condensable gas does not accumulate in the heat exchanger 16, and the wetwell gas supply pipe 48 can cool the high-pressure steam in the wetwell gas phase 7, preventing overpressure damage to the reactor containment vessel 3. Since there is always a pressure difference between the wet well gas phase 7 and the outer well 32, the non-condensable gas in the heat exchanger 16 is efficiently vented to the inside of the outer well 32 by the gas vent pipe 22. Therefore, in the passive containment vessel cooling system 12 of the conventional second nuclear power plant, it is not necessary to provide a vent fan in the gas vent pipe 22.

[0078] (Figure 21: Explanation of flooded piping) In this conventional example, in order to flood the lower dry well, submersion piping 68 is provided, which penetrates the wall of the pedestal 61, has one end led into the lower dry well (pedestal cavity) 61a, and the other end opens into the suppression pool 6. A submersion valve 67 is provided in the portion of the submersion piping 68 inside the lower dry well 61a. Also, the submersion piping 68 is provided with a check valve 69. By providing the check valve 69, it is possible to prevent high-temperature water in the lower dry well 61a from flowing back into the suppression pool 6 even if the pressure inside the dry well 4 rises. The submersion piping 68 is installed at positions midway between the LOCA vent pipes 8 so as not to overlap with them, for example, ten in total (see FIG. 22).

[0079] The flood valve 67 can be a melting valve. A melting valve is melted by the heat of the debris and opens. However, any valve that does not require a power source for operation during SBO can be used in addition to the melting valve. For example, an explosive valve can be used because it uses explosives for operation. A high temperature signal that detects the heat of the debris is used for operation. A spring-loaded valve can also be used because it uses the force of a spring for operation. The wire that holds the spring is melted by the heat of the debris and opens. Furthermore, to improve reliability, for example, five of the ten flood valves may be explosive valves and the remaining five may be spring-loaded valves. Also, five may be melting valves and the remaining five may be explosive valves. Alternatively, at least two or more types of melting valves, explosive valves, and spring-loaded valves may be used together.

[0080] (Figure 24: Explanation of PCCS heat exchange. In the case of a conventional second nuclear power plant) Fig. 24 is an elevational sectional view showing an example of a heat exchanger in the passive containment cooling system in the conventional nuclear power plant No. 2. The structure of the heat exchanger 16 in the passive containment cooling system 12 in the conventional nuclear power plant No. 2 will be described using an example of a horizontal heat exchanger with reference to Fig. 24.

[0081] In Figure 24, the outlet plenum 18 is provided below the inlet plenum 17. A large number of U-shaped heat transfer tubes 19 are connected to a tube sheet 23, and the straight tube portions of the heat transfer tubes 19 are installed horizontally. For simplification, only two tubes are shown in Figure 24. The outside of the heat transfer tubes 19 is filled with cooling water 14 (see Figure 21). The inlets of the heat transfer tubes 19 open to the inlet plenum 17. The outlets of the heat transfer tubes 19 open to the outlet plenum 18.

[0082] A wetwell gas supply pipe 48 is connected to the inlet plenum 17, and supplies a mixed gas of nitrogen, hydrogen, oxygen, water vapor, etc. in the wetwell gas phase section 7 to the inlet plenum 17. This mixed gas is led into the heat transfer tube 19, and the water vapor is condensed to become condensed water, which flows out from the outlet of the heat transfer tube 19 into the outlet plenum 18 and accumulates in the lower part of the outlet plenum 18.

[0083] A condensed water return pipe 21 is connected to the bottom of the outlet plenum 18, and the condensed water in the outlet plenum 18 is returned to the inside of the suppression pool 6 by gravity. In addition, a gas vent pipe 22 is connected to the top of the outlet plenum 18. Non-condensable gases such as nitrogen, hydrogen, and oxygen that do not condense in the heat transfer tubes 19 are discharged from the heat transfer tubes 19 and accumulate in the top of the outlet plenum 18.

[0084] The tip of the gas vent pipe 22 is immersed in the pool water of the scrubbing pool 33 in the external well 32. Non-condensable gas in the outlet plenum 18 passes through the gas vent pipe 22, pushes down the pool water of the scrubbing pool 33, and is vented into the water, and then moves into the space 33b between the top cover 33a and the scrubbing water 33c (see FIG. 23). The gas then moves into the interior of the external well 32 through the first outlet pipe 33d, the metal fiber filter 34, and the second outlet pipe 34a (see FIG. 23).

[0085] The shape of the heat transfer tubes 19 is not limited to a U-shape. There are also structures in which heat transfer tubes 19 with a straight tube section in the vertical direction are installed vertically (see Figure 25). The inlet plenum 17 is always located above the outlet plenum 18. This allows the condensed water that condenses inside the heat transfer tubes 19 to be guided to the outlet plenum 18 by gravity. The advantages of the horizontal type are excellent earthquake resistance and the ability to make effective use of the cooling water 14. On the other hand, the advantage of the vertical type is that the condensed water can be easily drained.

[0086] (Figure 25: Explanation of PCCS for conventional third-generation nuclear power plants) The configuration of the PCCS of the conventional third nuclear plant will be described with reference to Fig. 25. In the conventional third nuclear plant, a gravity-driven cooling system (GDCS) pool 28 is installed inside the upper dry well 4c. The height from the top slab 4a to the top end of the diaphragm floor 5b is about 6m. The inner diameter of the containment vessel 3 is about 36m. This configuration is adopted in the ESBWR (see Fig. 1 of Patent Document 2).

[0087] A cooling water pool 13 is provided on the top of the reactor containment vessel 3, and stores cooling water 14 inside. The cooling water pool 13 is of a pool type. A heat exchanger 16 is installed inside the cooling water pool 13. The heat exchanger 16 is of a vertical type. The heat exchanger 16 is installed so as to be completely submerged in the cooling water 14 during normal operation.

[0088] The heat exchanger 16 has an inlet plenum 17, an outlet plenum 18, and heat transfer tubes 19. The heat transfer tubes 19 are of a vertical type with their straight tube sections installed vertically. There are two inlet plenums 17, outlet plenums 18, and heat transfer tubes 19, and a pair of them constitute one heat exchanger 16. In many cases, multiple heat exchangers 16 are installed, but only one is shown in Figure 25. The heat exchanger 16 as a whole has a cooling capacity to condense steam at least equivalent to the decay heat.

[0089] The inlet plenum 17 of the heat exchanger 16 is connected to a drywell gas supply pipe 20. The drywell gas supply pipe 20 is made up of two pipes connected to the two inlet plenums 17 that are joined together along the way to form one pipe, which penetrates the top slab 4a and opens at the other end to the upper drywell 4c. The drywell gas supply pipe 20 does not have an isolation valve and is always open to the upper drywell 4c.

[0090] In addition, a condensate return pipe 21 is connected to the lower part of the outlet plenum 18 of the heat exchanger 16. The condensate return pipe 21 is formed by two pipes connected to the two outlet plenums 18, which each penetrate the top slab 4a and enter the upper dry well 4c, then joining to form one pipe. After joining together, the condensate return pipe 21 is connected to the PCCS drain tank 24.

[0091] The PCCS drain tank 24 is installed on the diaphragm floor 5b. Inside the PCCS drain tank 24, gas (mainly nitrogen gas) with the same composition as the gas in the upper dry well 4c is stored, but no water is stored. In addition, a debris protection lid 25 is provided on the top of the PCCS drain tank 24. The purpose of the debris protection lid 25 is to prevent as much as possible the intrusion of debris such as insulation that is scattered in the event of a loss of coolant accident from entering the PCCS drain tank 24.

[0092] The RPV water injection pipe 26 is connected to the lower part of the PCCS drain tank 24. The RPV water injection pipe 26 descends below the diaphragm floor 5b and the other end is connected to the RPV 2 in the lower region (see FIG. 18). The RPV water injection pipe 26 is provided with an RPV water injection isolation valve 27. The RPV water injection isolation valve 27 is normally closed. The RPV water injection isolation valve 27 opens when a loss of coolant accident occurs, the reactor water level drops, and a low reactor water level signal is generated. Therefore, even if a loss of coolant accident occurs, the RPV water injection isolation valve 27 does not open immediately, but opens with a time delay until a low reactor water level signal is generated. On the other hand, almost at the same time as the loss of coolant accident occurs, a large amount of steam generated from the break flow flows into the inside of the PCCS heat exchanger 16 from the drywell steam supply pipe 20, and a large amount of condensed water accumulates in the outlet plenum 18. If this condensed water is not discharged, the PCCS will lose its function. To avoid this, the PCCS drain tank 24 is required. Without the PCCS drain tank 24, the condensate cannot be discharged and will accumulate inside the PCCS heat exchanger 16, causing the PCCS heat exchanger to lose its function as it will be unable to draw in the steam in the upper dry well 4c.

[0093] A certain gap is provided between the debris protection cover 25 and the PCCS drain tank 24 so that the debris protection cover 25 does not seal the PCCS drain tank 24. This gap is necessary to allow the gas in the PCCS drain tank 24 to be discharged from the gap when a large amount of steam generated in the upper dry well 4c in the event of a loss of coolant accident is condensed in the PCCS heat exchanger 16 and flows into the PCCS drain tank 24 through the condensate water return pipe 21. If there is no gap and the PCCS drain tank 24 is sealed, the flow of condensate into the PCCS drain tank 24 is hindered. The gas in the PCCS drain tank 24 has nowhere to go and flows back through the condensate water return pipe 21 into the inside of the PCCS heat exchanger 16, hindering the steam condensation function of the PCCS heat exchanger. To avoid this, a certain gap is required between the debris protection cover 25 and the PCCS drain tank 24 so that the debris protection cover 25 does not seal the PCCS drain tank 24. This is also necessary because immediately after a loss of coolant accident, the RPV water injection isolation valve 27 is closed until a low reactor water level signal is generated. A certain gap is provided between the debris protection cover 25 and the PCCS drain tank 24, so the pressure in the upper dry well 4c acts through this gap on the surface of the condensed water that has flowed into the PCCS drain tank 24. In addition, when the amount of condensed water increases and the water level in the PCCS drain tank 24 rises, the condensed water overflows from the above gap into the upper dry well 4c. As a result, the condensed water in the PCCS drain tank 24 cannot rise above the height of the above gap.

[0094] The water level in the PCCS drain tank 24 is limited to about 3 m because the height of the gap is the upper limit. With a head pressure of only 3 m, it is difficult to inject the water in the PCCS drain tank 24 into the RPV 2 against the pressure in the RPV 2. In this state, in order to inject the condensed water in the PCCS drain tank 24 into the RPV 2 by opening the RPV water injection isolation valve 27, it is necessary to sufficiently lower the RPV water injection pipe 26 to the lower area 90 (see Figure 18) lower than the diaphragm floor 5b and then connect it to the RPV 2 in order to secure more head pressure. In order to lower the RPV water injection pipe 26 below the diaphragm floor 5b, it is necessary to secure a considerable piping space 29 between the LOCA vent pipe 8 and the RPV 2. Therefore, it is necessary to increase the outer diameter of the reactor containment vessel 3. Furthermore, since the inner diameter of the pedestal 61 also becomes larger, it is necessary to provide an RPV support 63 with a very special structure. The RPV support 63 needs to protrude significantly from the pedestal wall surface, which makes it difficult to ensure seismic strength. Similarly, the RPV water injection piping (not shown) of the GDCS pool 28 cannot inject water into the RPV 2 until it descends into the piping space 29. Therefore, the piping space 29 is essential in an ESBWR that uses a GDCS pool 28. This is an unavoidable problem for an ESBWR, which injects water from the upper dry well 4c into the RPV 2, which is also installed in the upper dry well 4c at approximately the same height, by gravity due to the difference in head. This is one of the problems with the ESBWR.

[0095] In addition, the GDCS pool 28 is installed inside the upper dry well 4c, and there is not enough space to install the PCCS drain tank 24, which blocks the opening of the LOCA vent pipe 8. The condensate return pipe 21 on the left side interferes with the GDCS pool 28. Therefore, in the final design of the ESBWR, the PCCS drain tank 24 was deleted, and a configuration was adopted in which the condensate return pipe 21 is connected to the inside of the GDCS pool 28. In addition, the RPV water injection pipe 26 and the RPV water injection isolation valve 27 were deleted.

[0096] In order to inject condensate from the PCCS heat exchanger 16 into the RPV 2 through the condensate return piping 21 in this manner, it was necessary to install the PCCS drain tank 24 inside the narrow upper dry well 4c, and further provide a piping space 29 for allowing the RPV water injection piping 26 to be lowered sufficiently to the lower region 90 (see Figure 18) below the diaphragm floor 5b, to enlarge the inner diameter of the reactor containment vessel 3, to enlarge the inner diameter of the pedestal 61, and to provide an RPV support 63 with a special structure. [Prior art documents] [Patent documents]

[0097] [Patent Document 1] JP 2016-14640 (US 2017 / 0162281 Al) [Patent Document 2] Japanese Patent Publication No. 2004-333357 (JP2004-333357A) [Patent Document 3] JP2007-232529A [Patent Document 4] JP2008-139023A Summary of the Invention [Problem to be solved by the invention]

[0098] The problem to be solved is that the passive containment cooling system 12 of a conventional nuclear power plant cannot cool the remaining debris in the reactor pressure vessel 2 in the event of a severe accident, and cannot prevent the upper lid 11a of the reactor containment vessel 3 from becoming hot and the seal material of the flange 74 from deteriorating, causing a large amount of hydrogen and radioactive material to leak outside the reactor containment vessel 3. Also, it cannot prevent the upper dry well 4c from becoming hot and causing overheating damage to the reactor containment vessel 3. Furthermore, in BWRs before ABWRs, it cannot prevent the water temperature in the pressure suppression pool from becoming stratified, causing high-pressure steam to be generated from the water at the surface of the water, causing early overpressure damage to the reactor containment vessel 3. [Means for solving the problem]

[0099] The present invention is characterized in that the condensed water of the passive containment vessel cooling system 12 is returned to the lower region 90 of the reactor containment vessel 3 and also injected into the upper region 91, and in the event of a severe accident, the condensed water is injected into the reactor pressure vessel 2 to cool the debris remaining in the reactor pressure vessel 2, suppressing a rise in temperature of the top lid 11a of the reactor containment vessel 3, preventing deterioration of the sealing material of the flange 74, and preventing overheating failure of the reactor containment vessel 3. Another feature is that the stratification of the water temperature in the suppression pool prevents early overpressure failure of the reactor containment vessel 3 due to high-pressure steam generated from the water on the water surface. Effect of the Invention

[0100] According to the embodiment of the present invention, even if a severe accident accompanied by core meltdown occurs due to a station blackout (SBO) or the like, it is possible to prevent overheating and early overpressure failure of the reactor containment vessel 3. Based on lessons learned from the Fukushima Daiichi accident, phenomena that may cause damage to the reactor containment vessel in the event of a severe accident can be suppressed, and the integrity of the reactor containment vessel can be maintained. [Brief description of the drawings]

[0101] [Figure 1] FIG. 1 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a first embodiment of the present invention. [Diagram 2] FIG. 2 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a second embodiment of the present invention. [Diagram 3] FIG. 3 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a third embodiment of the present invention. [Figure 4] FIG. 4 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a fourth embodiment of the present invention. [Diagram 5] FIG. 5 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a fifth embodiment of the present invention. [Figure 6]FIG. 6 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a sixth embodiment of the present invention. [Figure 7] FIG. 7 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a seventh embodiment of the present invention. [Figure 8] FIG. 8 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to an eighth embodiment of the present invention. [Figure 9] FIG. 9 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a ninth embodiment of the present invention. [Figure 10] FIG. 10 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a tenth embodiment of the present invention. [Figure 11] 11 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to an eleventh embodiment of the present invention. [Figure 12] FIG. 12 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a twelfth embodiment of the present invention. [Figure 13] FIG. 13 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a thirteenth embodiment of the present invention. [Figure 14] FIG. 14 is an elevational sectional view showing the configuration around a reactor containment vessel of a nuclear power plant according to a fourteenth embodiment of the present invention. [Figure 15] FIG. 15 is an elevational sectional view showing the configuration around the reactor containment vessel of a first conventional nuclear power plant (ABWR). [Figure 16] FIG. 16 is a plan view showing the configuration of the reactor containment vessel and its surroundings in a first conventional nuclear power plant (ABWR). [Figure 17] FIG. 17 is an elevational sectional view showing the configuration around the top cover of the reactor containment vessel of a conventional first nuclear power plant (ABWR). [Figure 18] FIG. 18 is a cross-sectional elevation view showing the internal region of the reactor containment vessel of a first conventional nuclear power plant (ABWR). [Figure 19]FIG. 19 is an elevational sectional view showing a reactor pressure vessel and reactor internals of a first conventional nuclear power plant (ABWR). [Figure 20] FIG. 20 is an elevational sectional view showing an example of a heat exchanger in a passive containment cooling system of a conventional first nuclear power plant (ABWR). [Figure 21] FIG. 21 is an elevational sectional view showing the configuration around the reactor containment vessel of the second conventional nuclear power plant (SATOR). [Figure 22] FIG. 22 is a plan view showing the configuration around the reactor containment vessel of the second conventional nuclear power plant (SATOR). [Figure 23] FIG. 23 is an elevational sectional view showing the configuration of a scrubbing pool of a conventional second nuclear power plant (SATOR). [Figure 24] FIG. 24 is an elevational sectional view showing an example of a heat exchanger of a passive containment cooling system of a conventional second nuclear power plant (SATOR). [Diagram 25] FIG. 25 is an elevational sectional view showing an example of a heat exchanger in a passive containment cooling system of a conventional third nuclear power plant (ESBWR). [Figure 26] FIG. 26 is an elevational sectional view showing the configuration around the reactor containment vessel, which shows a configuration common to the embodiments of the nuclear power plant of the present invention. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0102] A nuclear power plant having a passive containment cooling system according to an embodiment of the present invention will be described below with reference to Figures 1 to 14. Here, parts that are the same as or similar to the above-mentioned conventional technology and parts that are the same as or similar to the following embodiments will be given the same reference numerals, and duplicated explanations will be omitted, and only the main parts will be described.

[0103] (Description of ABWR-Based Embodiments) An embodiment based on ABWR will be described below with reference to Figures 1 to 7. In the following described embodiments 1 to 7, the configurations of the reactor core 1, reactor pressure vessel 2, and reactor containment vessel 3 are the same as those of the first conventional nuclear power plant (ABWR) shown in Figures 15 and 16. The basic configuration of the passive containment cooling system 12 is also the same as that of the first conventional nuclear power plant (ABWR) shown in Figures 15 and 20. In the following description, only the main points of the differences from the ABWR will be described for the first to seventh embodiments, avoiding duplication.

[0104] The embodiment shown in Fig. 1 to Fig. 7 uses a reactor containment vessel called RCCV, but the type of reactor containment vessel is not limited to RCCV. It is universally applicable to all pressure suppression type reactor containment vessels that have a pressure suppression function using a pressure suppression pool. In addition, other materials such as SC structure and steel can also be used. EXAMPLES

[0105] 1 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings of a nuclear power plant according to a first embodiment of the present invention. Hereinafter, differences between the first embodiment and a conventional ABWR will be described.

[0106] (FIG. 1: Explanation of the configuration of the PCCS of the first embodiment) The configuration of the passive containment cooling system (PCCS) 12 in the first embodiment will be described with reference to FIG. 1. In the first embodiment, one end of the condensate return pipe 21 is installed in the suppression pool 6 while being submerged in water, as an example. The condensate return pipe 21 is provided with a condensate return pipe isolation valve 21a. The condensate return pipe isolation valve 21a is open during normal operation. In the present embodiment, one end of the condensate return pipe 21 is led to the suppression pool 6, as an example, but is not limited to this example. There are also examples where the condensate return pipe 21 is led to a core catcher (not shown) below the lower dry well 61a or inside the LOCA vent pipe 8. Whether the condensate is led to the suppression pool 6 or inside the LOCA vent pipe 8, the condensate is finally returned from the melting valve 64 to the lower dry well 61a. In this way, one end of the condensate return pipe 21 is configured to return the condensate to the lower region 90 (see FIG. 18) below the diaphragm floor 5b of the containment vessel 3.

[0107] Furthermore, in this embodiment, a reactor pressure vessel water injection pipe 51 is provided. One end of the reactor pressure vessel water injection pipe 51 is connected between the outlet plenum 18 of the condensate return pipe 21 and the condensate return pipe isolation valve 21a. The other end of the reactor pressure vessel water injection pipe 51 is connected to the reactor pressure vessel 2. In this embodiment, an example is shown in which the other end of the reactor pressure vessel water injection pipe 51 is connected to the water supply sparger 50 in the reactor pressure vessel 2, but this example is not limiting. For example, it may be connected to the low pressure water injection system sparger 46 (see FIG. 19). The reactor pressure vessel water injection pipe 51 is provided with a reactor pressure vessel water injection pipe isolation valve 51a. The reactor pressure vessel water injection pipe isolation valve 51a is closed during normal operation.

[0108] Furthermore, the gas vent pipe 22 is provided with a gas vent fan 22a and a check valve 22b. In FIG. 1, the gas vent fan 22a is shown as being installed by providing a branch pipe from the gas vent pipe 22, but this is not the only possible example. The gas vent fan 22a can also be provided directly on the gas vent pipe 22. In this case, the check valve 22b does not need to be provided. In addition, in FIG. 1, the check valve 22b is shown as being a check valve, but this is not the only possible example. The check valve 22b can also be an electric valve or the like. A power source (not shown) dedicated to severe accidents is used to operate the gas vent fan 22a and the electric valve. The power source dedicated to severe accidents is a gas turbine generator, a battery, or the like, and is a diverse power source different from the emergency power source (diesel generator) for design basis accidents.

[0109] The first feature of this embodiment is that the condensate return pipe 21 is led into the suppression pool 6, and the reactor pressure vessel water injection pipe 51 is provided by branching from the middle of the condensate return pipe 21. The operation of this embodiment configured as above will be described. In the event of a loss of coolant accident, a large amount of steam is generated inside the upper dry well 4c from the rupture flow. The steam flows from the dry well gas supply pipe 20 into the heat exchanger 16 of the passive containment cooling system 12 under its own pressure, and is cooled and condensed by the heat transfer tube 19. The condensate flows into the outlet plenum 18. Furthermore, the condensate is returned to the suppression pool 6 by the condensate return pipe 21. Therefore, it is possible to prevent the passive containment cooling system 12 from losing function due to the condensate remaining in the heat exchanger 16. Since the suppression pool 6 functions as a PCCS drain tank in this way, it is not necessary to provide a PCCS drain tank 24 in the narrow upper dry well 4c as in the ESBWR (see FIG. 25). The operating method of the passive containment cooling system 12 in this manner, in which the condensate return pipe isolation valve 21a of the condensate return pipe 21 immersed in the suppression pool 6 is opened and the reactor pressure vessel water injection pipe isolation valve 51a is closed, is called the suppression pool drain mode.

[0110] On the other hand, if a core meltdown accident occurs due to an SBO or the like, the debris will melt through the bottom of the reactor pressure vessel 2 and fall to the floor of the lower drywell 61a. When the temperature in the lower drywell 61a rises due to the heat of the debris, the meltdown valve 64 opens and the water in the LOCA vent pipe 8 flows into the lower drywell 61a and floods the debris. The debris heats the water in the lower drywell 61a, generating a large amount of steam. The steam moves from the opening 66 of the lower drywell to the upper drywell 4c, and flows into the heat exchanger 16 of the passive containment cooling system 12 from the drywell gas supply pipe 20 under its own pressure, where it is cooled and condensed by the heat transfer pipe 19. This condensed water flows into the outlet plenum 18. Furthermore, the condensed water is returned to the suppression pool 6 by the condensed water return pipe 21. The water in the suppression pool flows into the lower drywell 61a again through the meltdown valve 64 via the LOCA vent pipe 8, and is used to cool the debris that has fallen to the floor of the lower drywell 61a. As a result, the cooling of the debris that has fallen to the lower D / W 61a is completely performed by the passive containment cooling system 12.

[0111] However, from the results of the Fukushima Daiichi accident, it has become clear that in the event of a severe accident, approximately 20% of the debris remains in the reactor pressure vessel 2. This remaining debris in the reactor is not submerged. If this state is left unattended, the temperature inside the reactor pressure vessel 2 will rise to approximately 1000°C, and due to heat radiation from the reactor pressure vessel 2, the temperature of the upper dry well 4c will exceed 200°C, and the temperature of the top cover 10a of the reactor containment vessel 3 will exceed 300°C, causing overheating and damage to the reactor containment vessel 3. Since the bottom of the reactor pressure vessel 2 is damaged, the pressure is reduced to approximately the pressure of the dry well 4. When this state is reached, signals indicating low reactor pressure vessel pressure and high radiation inside the reactor containment vessel are detected. When this signal is confirmed, the condensate return piping isolation valve 21a is closed and the reactor pressure vessel water injection piping isolation valve 51a is opened. This isolation valve switching may be performed manually or automatically by a signal. When the switching operation of this isolation valve is completed, the condensed water in the outlet plenum 18 is sprayed into the reactor pressure vessel 2 from the water supply sparger 50 in the reactor pressure vessel 2 by the reactor pressure vessel water injection piping 51. Since the height difference between the outlet plenum 18 and the water supply sparger 50 is about 10 m, the condensed water in the outlet plenum 18 can flow into the depressurized reactor pressure vessel 2 by gravity of about 10 m of head of water. This operating method of injecting condensed water into the reactor pressure vessel 2 by closing the condensed water return piping isolation valve 21a and opening the reactor pressure vessel water injection piping isolation valve 51a is called the reactor pressure vessel water injection mode of the passive containment cooling system 12. When the condensed water is sprayed into the reactor pressure vessel 2 from the water supply sparger 50, the temperature inside the reactor pressure vessel 2 drops to about 600°C or less. When the condensed water is sprayed from the water supply sparger 50, it descends the downcomer 2d (see FIG. 19) and flows from the diffuser 42 of the internal pump 40 to the lower plenum 2f at the bottom of the reactor pressure vessel 2. If the location of the melt penetration of the reactor pressure vessel 2 is an instrumentation pipe (not shown), the rupture may be higher than the bottom head 2g of the containment vessel 2, and the condensed water will accumulate up to that height and form a water level in the lower plenum 2f. The condensed water accumulated in the lower plenum 2f falls from the melt penetration location of the reactor pressure vessel 2 to the bottom of the lower dry well 61a and is used to cool the debris that has fallen to the floor of the lower dry well 61.If the condensed water is not sprayed, the temperature inside the reactor pressure vessel 2 will be maintained at a high temperature of about 1000°C due to the heat of the debris remaining in the reactor. Due to the heat radiation from the high-temperature reactor pressure vessel 2, the temperature of the upper dry well 4c may exceed 200°C and the temperature of the reactor containment vessel top lid 11a may exceed 300°C for seven days after the accident. This may cause the upper dry well 4c to overheat and break, deteriorating the seal material of the flange 74 of the reactor containment vessel top lid 11, causing a large amount of hydrogen and radioactive materials to leak to the operating floor. The Fukushima Daiichi accident has demonstrated that a large amount of leaked hydrogen may detonate on the operating floor. However, in this embodiment, the condensed water is sprayed from the feed water sparger 50 into the reactor pressure vessel 2, limiting the temperature inside the reactor pressure vessel 2 to about 600°C or less for seven days after the accident, so that the temperature of the upper dry well 4c is maintained at about 170°C or less and the temperature of the reactor containment vessel top lid 11a is maintained at about 200°C or less. This prevents the upper dry well 4c from being damaged by overheating, maintains the integrity of the seal material of the flange 74 of the reactor pressure vessel top head 11, and prevents a large amount of hydrogen and radioactive materials from leaking from the flange 74. In this way, the reactor pressure vessel water injection mode of the passive containment cooling system 12 is extremely effective in preventing the reactor containment vessel 3 from being damaged by overheating.

[0112] The present embodiment configured as described above can satisfy the configuration requirement of "cooling function of the molten core in the reactor (for example, low pressure water injection equipment into the reactor)" required by the Nuclear Regulation Authority (NRA) of Japan in the commentary of Article 42 of the New Regulatory Standards. Moreover, it can be satisfied by the passive containment cooling system 12 that does not use pumps or their power sources. This requirement can also be satisfied by an isolation condenser (IC), but since the isolation condenser is a core melt prevention equipment, its use is not permitted after the occurrence of a core meltdown in terms of defense in depth. If the isolation condenser is functioning, the core meltdown will not occur. The occurrence of a core meltdown accident means that the isolation condenser has broken down for some reason and cannot be used. This is what actually happened in the Fukushima Daiichi accident. Therefore, the isolation condenser cannot cool the molten core in the reactor after the occurrence of a core meltdown accident. In the case of the Fukushima Daiichi accident, the isolation condenser did not function at all after the occurrence of a core meltdown.

[0113] The second feature of this embodiment is that the gas vent pipe 22 is provided with a gas vent fan 22a. In this embodiment, the gas vent pipe 22 is led to the suppression pool 6. In the event of a severe accident, a large amount of radioactive material migrates from the safety relief valve 72 (see FIG. 15), the gas vent pipe 22, and the LOCA vent pipe 8 into the water of the suppression pool 6 and is present after being scrubbed. The amount is about 25% of the total radioactive material present in the core fuel. In other words, it is equivalent to substantially about 25% of the debris being present in the suppression pool 6. Therefore, the water in the suppression pool 6 starts to boil due to decay heat released by the radioactive material. The water in the suppression pool 6 boils and high-pressure steam flows into the wetwell gas phase 7, causing the pressure in the wetwell gas phase 7 to rise. The pressure in the drywell 4 is relatively low because it is cooled by the passive containment cooling system 12. Therefore, the pressure of the wetwell gas phase 7 becomes higher than the pressure of the drywell 4, and the nitrogen, hydrogen, and steam in the wetwell gas phase 7 become a mixed gas and flow into the drywell 4 through the vacuum breaker valve 9. This mixed gas flows into the heat exchanger 16 of the passive containment cooling system 12 through the drywell gas supply pipe 20 under its own pressure. The steam is condensed by the heat transfer tube 19. However, since nitrogen and hydrogen are non-condensable gases, they do not condense and remain inside the heat transfer tube 19. The non-condensable gas in the heat transfer tube 19 is originally configured to be vented into the suppression pool 6 through the gas vent tube 22 from the outlet plenum 18. However, when the water in the suppression pool 6 is boiling, the pressure of the wetwell gas phase 7 is higher than the pressure of the drywell 4, so the non-condensable gas in the heat transfer tube 19 cannot be vented into the suppression pool 6. Therefore, the passive containment cooling system 12 cannot draw in the gas in the drywell 4 and loses its cooling function. In other words, the passive containment cooling system 12 loses function when the water in the suppression pool 6 starts to boil. This phenomenon occurs almost simultaneously with the start of boiling of the water in the suppression pool 6. However, in this embodiment, a gas vent fan 22a is provided in the gas vent pipe 22 so that the non-condensable gas in the heat transfer tube 19 can be forcibly vented.This allows the non-condensable gas in the heat exchanger 16 to be vented into the suppression pool 6, enabling the passive containment cooling system 12 to maintain its cooling function. The passive containment cooling system 12 of this embodiment configured in this way can prevent overpressure damage to the reactor containment vessel 3 for a long period of time simply by replenishing the cooling water 14. The operating method in which the gas vent fan 22a is operated in this way is called the gas vent fan mode of the passive containment cooling system 12. The gas vent fan mode can be used in combination with other operating modes. EXAMPLES

[0114] 2 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings of a nuclear power plant according to a second embodiment of the present invention. Hereinafter, the points in which the second embodiment differs from a conventional ABWR will be described.

[0115] (FIG. 2: Explanation of the configuration of the PCCS in the second embodiment) The configuration of the passive containment cooling system (PCCS) 12 in the second embodiment will be described with reference to Fig. 2. In the second embodiment, a backflow prevention device 52 is further provided in the middle of the condensed water return pipe 21, and the end of the condensed water return pipe 21 is connected to a wetwell sparger 53 in the wetwell gas phase section 7. The other configurations of the passive containment cooling system (PCCS) 12 in the second embodiment are the same as those in the first embodiment. The wetwell sparger 53 can also use a sparger of the residual heat removal system (RHR) (not shown).

[0116] In FIG. 2, the backflow prevention device 52 is shown as a water seal tank, but is not limited to this example. The backflow prevention device 52 can be a U-shaped water seal or a check valve. In the case of the water seal tank 52, water is stored inside, and the condensate return pipe 21 is immersed in the water and opens halfway. In addition, a continued part of the condensate return pipe 21 is connected again to the part above the water level of the water seal tank 52, and the end is connected to the wetwell sparger 53 of the wetwell gas phase part 7. In the case of the U-shaped water seal, water is also stored inside. The purpose of the backflow prevention device 52 is to prevent the gas in the wetwell gas phase part 7 from flowing back through the condensate return pipe 21 and flowing into the heat exchanger 16 of the passive containment cooling system 12, causing the passive containment cooling system 12 to lose its cooling function. In the event of an SBO, reactor steam is released from the safety relief valve 72 (see FIG. 15) into the suppression pool 6. After a long time, high-pressure steam is generated from the water near the water surface of the suppression pool 6, and may flow back through the condensate return pipe 21 together with the non-condensable gas in the wetwell gas phase 7 and enter the heat exchanger 16 of the passive containment cooling system 12. The backflow prevention device 52 can prevent this backflow. Also, in the event of a severe accident, hydrogen generated in the reactor is released to the suppression pool 6 by the safety relief valve 72 and moves to the wetwell gas phase 7. As a result, hydrogen is added to the nitrogen that was originally present in the wetwell gas phase, and the pressure rises. There is a risk that this mixed gas of nitrogen and hydrogen will flow back through the condensate return pipe 21 and enter the heat exchanger 16 of the passive containment cooling system 12. The backflow prevention device 52 can prevent this backflow. It is considered that the depth of the water inside the water seal tank and U-shaped water seal for this backflow prevention should be about 0.25 m.

[0117] In addition, in the case of the water seal tank and the U-shaped water seal, in the event of a loss of coolant accident or a severe accident, it is possible to prevent gas containing radioactive materials in the drywell 4 from bypassing the gas vent pipe 22 and flowing directly into the wetwell gas phase part 7 from the condensate return pipe 21. In particular, since the large amount of steam in the drywell 4 immediately after a loss of coolant accident exceeds the cooling capacity of the heat exchanger 16, there is a risk that a large amount of steam that has not been condensed will flow directly into the wetwell gas phase part 7 from the condensate return pipe 21. This will hinder the pressure suppression function of the wetwell 5. In the case of the water seal tank and the U-shaped water seal, such bypass can be prevented by deepening the water depth inside. In this case, the water depth inside the water seal tank and the U-shaped water seal is made deeper than the submerged water depth of the gas vent pipe 22 in the pressure suppression pool 6 (see FIG. 1). This allows steam and gas to flow through the gas vent pipe 22 instead of the condensate return pipe 21.

[0118] In the second embodiment configured as above, in the event of a severe accident, the condensed water condensed by the heat exchanger 16 of the passive containment cooling system 12 can be sprayed from the wetwell sparger 53 to the inside of the wetwell gas phase 7 through the condensate return pipe 21. In the event of a severe accident, a large amount of steam and high-temperature hydrogen is released from the reactor pressure vessel 2 into the suppression pool 6 by the relief valve 72 (see FIG. 15). This causes the temperature of the pool water in the suppression pool 6 to rise. The lessons learned from the Fukushima Daiichi accident revealed that at that time, the temperature of the pool water becomes stratified, and the temperature of the water near the water surface becomes higher more quickly, generating high-pressure steam. When debris melts and penetrates the bottom of the reactor pressure vessel 2 and falls to the bottom of the lower drywell 61a, the melting valve 64 opens and water flows into the lower drywell 61a from the LOCA vent pipe 8, flooding the debris. Then, a large amount of steam is generated by the heat of the debris, and the drywell 4 becomes pressurized. The steam and nitrogen gas in the drywell 4 are released into the suppression pool 6 through the LOCA vent pipe 8 and the gas vent pipe 22 of the passive containment cooling system 12. The steam condenses with the water in the suppression pool 6, further increasing the temperature of the pool water. The nitrogen does not condense but moves to the wetwell gas phase 7. In this way, in the event of a severe accident, the wetwell gas phase 7 contains the nitrogen gas in the drywell 4 and a large amount of hydrogen generated by the metal-water reaction that has migrated from the safety relief valve 73 (see Figure 15), so the pressure in the wetwell gas phase 7 is already very high and is on the verge of overpressure failure. If the water in the suppression pool 6 becomes stratified, high-pressure steam will be generated from the water near the water surface at an earlier timing. This will cause the pressure in the wetwell gas phase 7 to increase more quickly, and the timing of overpressure failure of the reactor containment vessel 3 will become earlier. In order to prevent overpressure failure, the Nuclear Regulation Authority of Japan (NRA) requires the use of a filter vent system. A filter vent system is a dangerous device that releases radioactive rare gases and organic iodines directly into the environment. It is a dangerous device that prioritizes preventing damage to the reactor containment vessel 3 and preventing land contamination. Therefore, it is necessary to evacuate all residents in the surrounding area before activating the filter vent. It takes time to evacuate residents.However, due to the stratification of the suppression pool 6, the timing of activating the filter vent was premature, and there was a risk that radioactive rare gases and organic iodines would be released into the environment before the evacuation of residents was completed, resulting in exposure to residents. In the event of an SBO, local governments have formulated emergency disaster prevention plans in which residents begin to prepare for preventive evacuation at the time of the SBO occurrence, and furthermore, when the cooling function is lost, the evacuation of residents begins. However, in the worst-case scenario, it takes about 2 hours from the occurrence of an SBO until the core meltdown begins, and about 5 hours until the meltdown and penetration of the reactor pressure vessel 2. Furthermore, if the water in the suppression pool 6 becomes stratified, the timing until the reactor containment vessel 3 is overpressurized and breaks is about 20:00 to about 55 hours in the case of an ABWR. If an SBO occurs at night due to a massive earthquake, whether the evacuation of residents left behind in collapsed houses can be completed within about 20 hours to about 55 hours using destroyed roads and unlit traffic lights depends on the content of the emergency disaster prevention plan. About 20 hours is when about 20% of the debris remains in the reactor. This assumes that there will be no remaining debris inside the reactor for approximately 55 hours.

[0119] However, according to this embodiment, it is possible to spray condensed water generated in the heat exchanger 16 of the passive containment cooling system 12 to the wetwell gas phase 7. The temperature of the sprayed water is about 60°C at the beginning of the accident because the temperature of the pool water 14 of the passive containment cooling system 12 is low. After that, when the temperature of the pool water 14 reaches about 100°C, the temperature of the sprayed water becomes constant at about 130°C. On the other hand, if the suppression pool becomes stratified, the water temperature near the water surface reaches about 180°C. However, in this embodiment, the high-temperature steam generated from the water in the suppression pool 6 can be condensed by the sprayed water. In addition, since the sprayed water falls on the water surface of the suppression pool 6 that is at the highest temperature, it is possible to prevent the water in the suppression pool 6 from becoming stratified and to maintain the temperature at about 130°C, which is the same as the temperature of the sprayed water. In addition, the decay heat generated from the debris, including the decay heat from the debris remaining in the reactor and the decay heat from the debris that has fallen to the floor of the lower dry well 61a, can be cooled by the passive containment cooling system 12. For this reason, it becomes possible to postpone the timing of overpressure failure of the reactor containment vessel 3 to three days or later. By replenishing the pool water 14, it becomes possible to further postpone the timing of overpressure failure of the reactor containment vessel 3 and prevent overpressure failure itself.

[0120] This operating method of closing the reactor pressure vessel water injection isolation valve 51a and opening the condensate return pipe isolation valve 21a to spray condensate into the wetwell gas phase 7 is called the wetwell spray mode of the passive containment cooling system 12. A plurality of heat exchangers 16 are installed in the passive containment cooling system 12. For example, if four heat exchangers 16 are installed, it is conceivable to operate two of the heat exchangers 16 in the wetwell spray mode and the remaining two in the reactor pressure vessel water injection mode. By doing so, it becomes possible to prevent overpressure failure and overtemperature failure of the reactor containment vessel 3. EXAMPLES

[0121] 3 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings of a nuclear power plant according to a third embodiment of the present invention. Hereinafter, differences between the third embodiment and a conventional ABWR will be described.

[0122] (Description of the configuration of PCCS in the third embodiment) The configuration of the passive containment cooling system (PCCS) 12 in the third embodiment will be described with reference to Fig. 3. In the third embodiment, the reactor pressure vessel water injection piping 51 is connected to a high pressure core water injection system sparger 47 (see Fig. 19) in the shroud head 2b in the reactor pressure vessel 2. The other configurations of the passive containment cooling system (PCCS) 12 in the third embodiment are the same as those in the second embodiment.

[0123] In the third embodiment configured as described above, it becomes possible to spray the condensed water of the passive containment cooling system 12 into the shroud head 2b (see FIG. 19) in the reactor pressure vessel 2. This makes it possible to spray the sprayed water directly onto the core 1 inside the shroud 2a. This makes it possible to spray the water directly onto the debris remaining in the reactor, and further makes it possible to reduce the temperature inside the reactor pressure vessel 2.

[0124] Furthermore, if the bottom head 2g of the reactor pressure vessel 2 is largely melted and penetrated during a severe accident, in the case of the second embodiment, the condensed water sprayed from the water supply sparger 50 descends the downcomer 2d, passes through the diffuser 42 of the internal pump 40, falls into the lower plenum 2f of the reactor pressure vessel 2, and then falls into the lower dry well 61a from the broken part of the bottom head 2g of the reactor pressure vessel (see FIG. 19). Thus, in the case of the second embodiment, the condensed water flows out of the reactor pressure vessel 2 without cooling the debris remaining in the reactor. Therefore, if the bottom head 2g of the reactor pressure vessel 2 is largely melted and penetrated, there is a risk that a sufficient cooling effect cannot be obtained.

[0125] However, in the third embodiment, even if the bottom head 2g of the reactor pressure vessel 2 has been significantly melted and penetrated, water can be directly sprayed from the high-pressure core injection system sparger 47 onto the remaining debris in the reactor to cool it, making it possible to obtain a sufficient cooling effect. EXAMPLES

[0126] 4 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings of a nuclear power plant according to a fourth embodiment of the present invention. Hereinafter, differences between the fourth embodiment and a conventional ABWR will be described.

[0127] (Description of the configuration of PCCS in the fourth embodiment) The configuration of the passive containment cooling system (PCCS) 12 in the fourth embodiment will be described with reference to FIG. 4. In the fourth embodiment, the reactor pressure vessel water injection pipe 51 is connected to the high pressure core injection system water injection pipe 95 of the high pressure core injection system 92. The connection position is between the high pressure core injection system outer isolation valve 96 and the reactor pressure vessel 2. FIG. 4 shows an example in which the high pressure core injection system is connected between the high pressure core injection system inner isolation valve 97 and the reactor pressure vessel 2, but the connection position is not limited to this example. For example, the high pressure core injection system outer isolation valve 96 may be connected between the reactor pressure vessel 3. In this case, the reactor pressure vessel water injection pipe 51 does not need to penetrate the reactor containment vessel 3. In this embodiment, the condensed water can be sprayed from the high pressure core injection system sparger 47 (see FIG. 19) in the shroud head 2b in the reactor pressure vessel 2 by utilizing the high pressure core injection system water injection pipe 95 of the high pressure core injection system 92 that already exists. Other configurations of the passive containment cooling system (PCCS) 12 in the fourth embodiment are the same as those in the second embodiment.

[0128] The ABWR is provided with two high-pressure core cooling systems 92. This was implemented by the inventor of the present invention in order to reduce the occurrence frequency of an automatic depressurization system (ADS) (not shown) and to reduce the occurrence frequency of a high-pressure scenario called a TQUX sequence in a probabilistic safety assessment (PRA). On the other hand, a plurality of heat exchangers 16, for example, four, are provided in the passive containment cooling system 12. In this case, it is considered that the reactor pressure vessel water injection pipes 51 of the two heat exchangers 16 are connected to the high-pressure core water injection system water injection pipes 95 of the high-pressure core water injection system 92 according to the fourth embodiment. The remaining two heat exchangers 16 may be configured in such a way that the reactor pressure vessel water injection pipes 51 and the reactor pressure vessel water injection pipe isolation valves 51a are deleted from the second embodiment. The former two heat exchangers 16 are operated in a reactor pressure vessel water injection mode to prevent overheating damage of the reactor containment vessel 3. On the other hand, the latter two heat exchangers 16 can be operated in wetwell spray mode to prevent overpressure failure of the reactor pressure vessel 3. In this way, both overtemperature failure and overpressure failure of the reactor pressure vessel 3 can be prevented by the passive containment cooling system 12. EXAMPLES

[0129] 5 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings of a nuclear power plant according to a fifth embodiment of the present invention. Hereinafter, the differences between the fifth embodiment and a conventional ABWR will be described.

[0130] (Description of the configuration of PCCS in the fifth embodiment) The configuration of a passive containment cooling system (PCCS) 12 in a fifth embodiment will be described with reference to Fig. 5. In the fifth embodiment, a drywell gas supply piping isolation valve 20a is provided in the drywell gas supply piping 20. The drywell gas supply piping isolation valve 20a is shown in a closed state in Fig. 5, but is open during normal operation. However, the drywell gas supply piping 20a is closed during normal operation and may be opened after an accident.

[0131] In this embodiment, a reactor pressure vessel (RPV) gas supply pipe 54 is further provided so as to be connected to the drywell gas supply pipe 20. One end of the reactor pressure vessel (RPV) gas supply pipe 54 is connected to the drywell gas supply pipe 20 between the inlet header 17 of the heat exchanger 16 and the drywell gas supply pipe isolation valve 20a. The reactor pressure vessel (RPV) gas supply pipe 54 penetrates the reactor containment vessel 3, and the other end is connected to the steam phase 2s (see FIG. 19) of the reactor pressure vessel 2. The reactor pressure vessel gas supply pipe 54 is provided with a reactor pressure vessel gas supply pipe isolation valve 54a. The reactor pressure vessel gas supply pipe isolation valve 54a is shown in the figure in an open state, but is closed during normal operation. The reactor pressure vessel gas supply pipe isolation valve 54a can be an electric valve, an air-operated valve, a blast valve, or the like. In this embodiment, an example is shown in which one end of the reactor pressure vessel gas supply pipe 54 is connected to the drywell gas supply pipe 20, but the present invention is not limited to this example. One end of the reactor pressure vessel gas supply piping 54 may also be connected directly to the inlet plenum 17 of the heat exchanger 16 .

[0132] In this embodiment, a pressure equalizing valve 55 is further provided. The pressure equalizing valve 55 is connected to the steam phase 2s (see FIG. 19) of the reactor pressure vessel 2 by a stub pipe 55a. The pressure equalizing valve 55 is shown in an open state in FIG. 5, but is closed during normal operation. In FIG. 5, the stub pipe 55a is shown to be directly connected to the reactor pressure vessel 2, but is not limited thereto. For example, the stub pipe 55a may be connected to the main steam pipe 71 (see FIG. 15). The stub pipe 55a is configured to release the gas of the steam phase 2s of the reactor pressure vessel 2 from the pressure equalizing valve 55 to the upper dry well 4c to equalize the pressures of the reactor pressure vessel 2 and the dry well 4c. The pressure equalizing valve 55 may be an electrically operated valve, an air operated valve, a blast valve, or the like. The configuration of the pressure equalizing valve 55 is similar to that of the pressure reducing valve of the ESBWR. However, the pressure equalizing valve 55 is configured to operate only after the reactor pressure vessel 2 is in a low pressure state. On the other hand, the pressure reducing valve of the ESBWR operates when the reactor pressure vessel 2 is in a high pressure state. If the pressure reducing valve operates from a high pressure state during a severe accident, a large amount of radioactive material is released into the drywell 4 and deposited on the walls and floor of the drywell 4, and the decay heat is continuously released, so that the drywell 4 may overheat and be damaged. This is one of the problems of the ESBWR using the pressure reducing valve. On the other hand, the pressure equalizing valve 44 does not operate when the reactor pressure vessel 2 is in a high pressure state, so that a large amount of radioactive material in the reactor pressure vessel 2 is released from the safety relief valve 72 (see FIG. 15) to the suppression pool 6. Therefore, even if the pressure equalizing valve 55 is opened after the reactor pressure vessel 2 is depressurized, a large amount of radioactive material is not released into the drywell 4 and the drywell 4 does not overheat and be damaged. This is the difference between the pressure equalizing valve 55 and the pressure reducing valve of the ESBWR. The other configurations of the passive containment cooling system (PCCS) 12 in the fifth embodiment are the same as those in the fourth embodiment.

[0133] The operation of the fifth embodiment configured as above during a severe accident will be described below. When a core meltdown occurs due to an SBO or the like, the bottom of the reactor pressure vessel 2 is damaged, and debris falls to the bottom of the lower drywell 61a. The heat of the debris opens the meltdown valve 64, and water from the pressure suppression pool 6 flows into the lower D / W 61a from the LOCA vent pipe 8, flooding the debris that has fallen to the bottom of the lower drywell 61a. A large amount of steam is generated from the water flooded by the heat of the debris. The steam moves from the opening 66 to the upper drywell 4c and forms a mixed gas with nitrogen, hydrogen, etc. This mixed gas flows into the heat exchanger 16 through the drywell gas supply pipe 20 under its own pressure. The steam is condensed by the heat transfer pipe 19, flows into the outlet plenum, and further passes through the condensed water return pipe 21 and is sprayed from the wetwell sparger 53 to the wetwell gas phase 7 (see FIG. 3). That is, the heat exchanger 16 is operated in the wetwell spray mode. Nitrogen and hydrogen are non-condensable gases, so they do not condense and are vented from the gas vent pipe 22 into the suppression pool 6. If this state continues for a while, almost all of the nitrogen and hydrogen in the drywell 4 will be vented into the suppression pool 6 and move to the wetwell gas phase 7. As a result, after about one hour, the inside of the drywell 4 will be occupied only by steam.

[0134] In this state, the bottom of the reactor pressure vessel 2 has already been damaged, so the pressure in the reactor pressure vessel 2 has dropped, and signals indicating low reactor pressure and high radiation in the reactor containment vessel are generated. When this state is reached, the condensate return pipe isolation valve 21a is closed, the reactor pressure vessel water injection pipe isolation valve 51a is opened, and the passive containment vessel cooling system 12 starts operating in the reactor pressure vessel water injection mode (see FIG. 3). In this embodiment shown in FIG. 5, the pressure equalization valve 55 and the reactor pressure vessel gas supply pipe isolation valve 54a are further opened, and the drywell gas supply pipe isolation valve 20a is closed. These valves may be switched manually or automatically in response to signals indicating low reactor pressure vessel pressure and high radiation in the reactor containment vessel.

[0135] When the valves are switched in this way, the gas in the drywell 4 flows into the reactor pressure vessel 2 through the rupture at the bottom of the reactor pressure vessel 2 and the pressure equalizing valve 55 because the drywell gas supply piping isolation valve 20a is closed, and flows into the inside of the reactor pressure vessel 2 through the reactor pressure vessel gas supply piping 54 due to its own pressure and into the inside of the heat exchanger 16. The steam is condensed by the heat transfer tube 19 and flows into the outlet plenum 18. Since the passive containment cooling system 12 has already started operating in the reactor pressure vessel injection mode, the condensed water in the outlet plenum 18 is sprayed into the inside of the reactor pressure vessel 2 through the reactor pressure vessel water injection piping 51 (see Figure 3).

[0136] In this embodiment, the steam in the drywell 4 first passes through the inside of the reactor pressure vessel 2 and then flows into the heat exchanger 16 of the passive containment cooling system 12. The flow rate is very large, about 10,000 m3 / h. The steam generated from the water that submerges the debris in the lower drywell 61a is saturated steam at about 130°C. The steam flows into the inside of the reactor pressure vessel 2 at an enormous flow rate and flows out to the passive containment cooling system 12. In the process, the debris remaining in the reactor pressure vessel 2 is cooled by the enormous air-cooling effect of the steam. The same air-cooling effect as cooling with a huge electric fan is obtained. As a result, the temperature inside the reactor pressure vessel 2 is further reduced to about 250°C to about 350°C. In this embodiment, the low-temperature steam flows in the steam dome 2e of the reactor pressure vessel 2 at an enormous flow rate, so that the temperature of the reactor pressure vessel head 2c drops, and the temperature of the top lid 11a of the reactor containment vessel 3 drops further to a temperature well below 200°C, so that it is possible to more reliably prevent the deterioration of the seal material of the flange 74 of the top lid 11 of the reactor containment vessel 3 and the leakage of a large amount of hydrogen and radioactive materials to the operating floor. This operating method in which the pressure equalizing valve 55 is opened, the reactor pressure vessel steam supply piping isolation valve 54a is opened, and the drywell gas supply piping isolation valve 20a is closed to introduce the steam in the drywell 4 into the reactor pressure vessel 2 and cool the inside of the reactor pressure vessel 2 with the air-cooling effect of the enormous steam flow rate is called the reactor pressure vessel steam cooling mode of the passive containment vessel cooling system 12. The reactor pressure vessel steam cooling mode can be used in combination with other operating modes.

[0137] The present embodiment configured as described above can satisfy the configuration requirements of "cooling function of the molten core in the reactor (for example, low pressure water injection equipment into the reactor)" required by the Nuclear Regulation Authority (NRA) of Japan in the commentary of Article 42 of the New Regulatory Standards. Moreover, it can be satisfied by the passive containment cooling system 12 that does not use a pump or its power source. The NRA refers to the cooling function of the molten core in the reactor, and the method is exemplified by low pressure water injection equipment, but is not limited to this. Therefore, the cooling function of the molten core in the reactor by the air cooling effect of steam according to this embodiment satisfies the requirements of the NRA. The present invention provides the cooling function of the molten core in the reactor at low cost and with high reliability by the passive air cooling function of the passive containment cooling system 12, without using the dynamic pump or power source of the low pressure water injection equipment, by an innovative method that the NRA could not have conceived. The energy source for its operation is the decay heat generated by the debris. Since it is no longer necessary to provide a low pressure water injection equipment for severe accidents and a power source for severe accidents, it is possible to achieve both cost reduction and reliability improvement at the same time. The inventor calls this design philosophy, which simultaneously achieves cost reduction and performance improvement, the Positive Cost Down Philosophy. In addition, the design philosophy that achieves this with the passive containment cooling system 12, which is an already existing technology, the Off the Shelf Philosophy. These are made possible by the newly created value created by the designer's invention.

[0138] In addition, in order for the steam in the drywell 4 to push down the water inside the LOCA vent pipe 8 and flow into the suppression pool 6 during a severe accident (called vent clearing), the pressure of the steam in the drywell 4 needs to be high. If the water depth of the suppression pool 6 is deep enough, the pressure required for vent clearing becomes large and becomes larger than the pressure difference between the reactor pressure vessel 2 and the drywell 4, and the steam in the drywell 4 can flow into the reactor pressure vessel 2 from the rupture at the bottom of the reactor pressure vessel 2 without providing the pressure equalizing valve 55. In other words, if the water depth of the suppression pool is deep enough, the pressure equalizing valve 55 is not essential. Conversely, if the water depth of the suppression pool 6 is not deep enough, the pressure required for vent clearing becomes small, so the steam in the drywell 4 vent-clears the LOCA vent pipe 8 and flows into the suppression pool 6. In that case, the steam in the drywell 4 flows into the suppression pool 6 without flowing into the reactor pressure vessel 2. In order to avoid this, the pressure equalizing valve 55 is effective. EXAMPLES

[0139] 6 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings in a nuclear power plant according to a sixth embodiment of the present invention. Hereinafter, the points in which the sixth embodiment differs from a conventional ABWR will be described.

[0140] (Description of the configuration of PCCS in the sixth embodiment) The configuration of the passive containment cooling system (PCCS) 12 in the sixth embodiment will be described with reference to FIG. 6. In the sixth embodiment, a drywell spray pipe 56 is provided connected to the condensate return pipe 21. The connection point of the drywell spray pipe 56 with the condensate return pipe 21 is located below the outlet plenum 18 of the heat exchanger 16 and above the connection point of the reactor pressure vessel water injection pipe 51 and the condensate return pipe 21. The drywell spray pipe 56 is provided with a drywell spray pipe isolation valve 56a. The other end of the drywell spray pipe 56 passes through the reactor containment vessel 3 and is connected to a drywell spray sparger 56b. The drywell spray sparger 56b can use a reactor containment vessel spray sparger of the residual heat removal system (RHR) (not shown). The other configurations of the passive containment cooling system (PCCS) 12 in the sixth embodiment are the same as those in the fifth embodiment.

[0141] In this embodiment configured as above, when the condensate return pipe isolation valve 21a is closed, the reactor pressure vessel water injection pipe isolation valve 51a is closed, and the drywell spray pipe isolation valve 56a is opened in the event of a severe accident, the condensate in the outlet plenum 18 of the heat exchanger 16 can be sprayed from the drywell spray sparger 56b into the upper drywell 4c through the drywell spray pipe 56. Such an operation method of the passive containment cooling system 12 is called the drywell spray mode. Since the height difference between the outlet plenum 18 and the drywell spray sparger 56b is about 5m or more, it is possible to spray the condensate into the drywell 4. The drywell spray mode cannot be used together with the reactor pressure vessel water injection mode and the wetwell spray mode in the same passive containment cooling system 12. It can be used together with the reactor pressure vessel steam cooling mode and the gas vent mode. However, it can be used together with all other operation modes between different passive containment cooling systems 12.

[0142] The operation and effect of this embodiment will be described below. Assume that the main steam pipe 71 (see FIG. 15) breaks inside the reactor containment vessel 3, and then a core meltdown accident occurs. If a huge earthquake occurs and the main steam pipe 71 breaks, and an SBO occurs at the same time, there is a risk of a core meltdown accident after the main steam pipe breaks. In that case, a large amount of hydrogen and radioactive material generated during the progress of the core meltdown is released from the break in the main steam pipe 71 into the upper dry well 4c. The hydrogen is vented into the pressure suppression pool 6 by the LOCA vent pipe 8 and the gas vent pipe 20, and moves to the wet well gas phase 7. On the other hand, although a part of the large amount of radioactive material released into the upper dry well 4c moves to the pressure suppression pool 6, most of it adheres to the wall and floor of the upper dry well 4c and continues to emit decay heat. The amount of radioactive material released into the dry well 4c is about 25% of the fission products that were present inside the core fuel. In the case of a normal severe accident, approximately 25% of these fission products are transferred to the pressure suppression pool 6 by the safety relief valve 72 (see Figure 15) and are cooled by the pool water. However, in the case of a main steam pipe rupture accident, approximately 25% of the fission products are released from the rupture in the main steam pipe 71, and most of them adhere to the inside of the upper dry well 4c, existing in the gas phase and not being cooled. This is a terrifying situation that exceeds that of the Fukushima Daiichi accident. If this condition is left unchecked, the temperature inside the upper dry well 4c will exceed 400°C and the reactor containment vessel 3 will overheat and be damaged.

[0143] On the other hand, a large amount of debris, exceeding 30%, remains in the reactor pressure vessel 2 without falling into the lower dry well 61a. Because heat is released from the broken portion of the main steam pipe 71 and cooled, more debris remains in the reactor than in a normal severe accident. However, after the bottom of the reactor pressure vessel 2 melts and breaks, the broken opening of the main steam pipe 71 has an effect equal to or greater than that of the pressure equalizing valve 55 because the main steam pipe 71 has a large diameter, and steam in the dry well 4 flows in from the broken opening at the bottom of the reactor pressure vessel 2 and flows out from the broken opening of the main steam pipe 71. The temperature in the reactor pressure vessel 2 falls below about 400°C due to the air-cooling effect of this steam flow, which cools the debris remaining in the reactor. In this state, there is little effect or need to cool the debris remaining in the reactor by using the reactor pressure vessel injection mode. On the other hand, the temperature in the upper dry well 4c will exceed 400°C if left unattended due to the decay heat of the attached radioactive material. This is a frightening state.

[0144] This situation also occurs in the case of a loss of coolant accident in a PWR. A large amount of fission products is released into the reactor containment vessel from the ruptured cold leg piping, and adheres to the walls and floor, continuing to release decay heat without being cooled. Even if a core catcher is installed to cool the debris, or in vessel retention is implemented to prevent melting and damage to the lower part of the reactor pressure vessel, it is almost meaningless. The reactor containment vessel will overheat and break due to the decay heat of the radioactive material that adheres to the inside of the reactor containment vessel. In particular, if in vessel retention is implemented, the reactor containment vessel will overheat and break due to the decay heat from 100% of the debris remaining in the reactor pressure vessel. Of course, there are designs in PWRs that deal with this phenomenon, such as directly cooling the outer surface of the reactor containment vessel with water and air cooling, or spraying containment vessel sprays specially designed for severe accidents. However, some PWRs are designed so that only a core catcher is needed after a severe accident, and do not cool the radioactive material attached to the gas phase of the reactor containment vessel. This is because the lessons learned from the Fukushima Daiichi accident have not yet been fully recognized, as models for analyzing physical phenomena during severe accidents that can handle heat release from the debris remaining in the reactor have not yet been developed for PWR cores. As a result, some PWRs are designed with the understanding that if all debris in the reactor falls into the core catcher in the event of a severe accident and the core catcher is flooded, then all that is left to do is deal with overpressure failure. This is frightening. The lesson learned from the Fukushima Daiichi accident was that in the event of a severe accident, the reactor containment vessel would fail due to overheating rather than melting through, but measures to this effect have not been fully implemented.

[0145] As described above, in the case of a severe accident after the breakage of the main steam pipe, about 30% or more of debris remains in the reactor, but steam flows from the breakage at the bottom of the RPV to the breakage of the main steam pipe 71, and the debris remaining in the reactor is cooled by the air-cooling effect of the steam. Meanwhile, a large amount of radioactive material is released from the breakage of the main steam pipe 71 into the upper dry well 4c, and adheres to the wall and floor surfaces, and continues to release decay heat without being cooled. If this state is left unattended, the temperature in the upper dry well 4c will exceed 400°C. However, in this embodiment, condensed water at about 130°C can be sprayed inside the upper dry well 4c by switching the passive containment cooling system 12 to the dry well spray mode. This makes it possible to sufficiently lower the temperature in the upper dry well 4c. EXAMPLES

[0146] 7 is an elevational sectional view showing the configuration of the reactor containment vessel and its surroundings of a nuclear power plant according to a seventh embodiment of the present invention. Hereinafter, the points in which the seventh embodiment differs from a conventional ABWR will be described.

[0147] (Description of the configuration of PCCS in the seventh embodiment) The configuration of the passive containment cooling system (PCCS) 12 in the seventh embodiment will be described with reference to Fig. 7. In the seventh embodiment, a backflow prevention device (water seal tank) 52 is provided in a portion of the condensate return pipe 21 below the outlet plenum 18 of the heat exchanger 16 and above the connection point between the drywell spray pipe 56 and the condensate return pipe 21. The other configurations of the passive containment cooling system (PCCS) 12 in the seventh embodiment are the same as those in the sixth embodiment.

[0148] In this embodiment, since the backflow of the drywell spray piping 56 can be prevented, the drywell spray piping isolation valve 56a can be opened during normal operation. In this state, even if a loss of coolant accident occurs, the gas in the upper drywell 4c does not flow back through the drywell spray piping 56. Most of the steam in the upper drywell 4c is mixed with the nitrogen in the upper drywell 4c and vented into the suppression pool 6 through the LOCA vent pipe 8 due to the large amount of steam in the blowdown process immediately after the LOCA. Some of the steam also flows into the passive containment cooling system 12, and the nitrogen is vented into the suppression pool 6 through the gas vent pipe 22. During this process, almost all of the nitrogen in the drywell 4 moves to the wetwell gas phase. When the blowdown process is completed and the amount of steam generated is stabilized, all of the steam in the upper drywell 4c flows into the heat exchanger 16 through the drywell gas supply piping 20 instead of the LOCA vent pipe 8. This is because the submergence depth of the gas vent pipe 22 in the suppression pool 6 is shallower than the submergence depth of the top of the horizontal vent pipe 8a (see Figure 1). The gas in the drywell 4 moves to the suppression pool 6 through a route with a shallower submergence depth. The steam is condensed by the heat transfer pipe 19, and the condensed water flows into the outlet plenum 18. The steam and non-condensable gases that were not condensed are vented to the suppression pool 6 by the gas vent pipe 22. The steam is condensed in the suppression pool 6. The non-condensable gases move to the wetwell gas phase 7. The condensed water that flows into the outlet plenum 18 passes by gravity through the condensed water return pipe 21 and the drywell spray pipe 56, and is sprayed from the drywell spray sparger 56b into the upper drywell 4c. The temperature of the condensed water immediately after the accident is low, about 60°C, because the temperature of the cooling water 14 is low. By spraying the low-temperature condensed water into the upper dry well 4c, the steam in the upper dry well 4c condenses, and the pressure in the reactor containment vessel 3 is rapidly reduced. In a conventional ABWR, when a loss of coolant accident occurs, the pressure in the reactor containment vessel 3 remains high until the residual heat removal system (RHR) (not shown), which has been used as a low-pressure water injection system (LPFL), is manually switched after 15 minutes to operate the containment vessel spray and spray the coolant.Furthermore, if the manual switchover of the containment vessel spray fails, there is a risk that the pressure in the containment vessel 3 will remain high thereafter. Considering the occurrence of a loss of coolant accident due to a huge earthquake, it is not easy to switch the LPFL, which is injecting water into the reactor as an ECCS, to the containment vessel spray within 15 minutes in a confused state. In the first place, it is not easy to accurately identify what kind of accident has occurred in 15 minutes. However, in this embodiment, since the drywell spray piping isolation valve 56b is normally open, when a loss of coolant accident occurs, cold and hot water is statically and automatically sprayed into the upper drywell 4c by the drywell spray mode of the passive containment vessel cooling system 12, so that the pressure in the containment vessel 3 naturally decreases. While the operators are confused and wondering what to do, the pressure in the containment vessel 3 naturally decreases. When a loss of coolant accident occurs, the passive containment vessel cooling system 12 in this embodiment can return the generated steam to low-temperature condensed water and spray it into the upper drywell 4c as a spray.

[0149] (Description of an embodiment based on SATOR) Hereinafter, an embodiment based on Sever Accident Tolerant and Optimized Rector (SATOR) will be described with reference to Fig. 8 to Fig. 14. In the following eighth to fourteenth embodiments, the configurations of the reactor core 1, reactor pressure vessel 2, reactor containment vessel 3, and outer well 32 are the same as those of the conventional second nuclear power plant (SATOR) shown in Fig. 21 and Fig. 22. The basic configuration of the passive containment cooling system 12 is also the same as that of the passive containment cooling system 12 of the conventional second nuclear power plant (SATOR) shown in Fig. 21 and Fig. 24. In the following description, only the main points of the differences from SATOR for the eighth to fourteenth embodiments will be described to avoid duplication.

[0150] The embodiment shown in Fig. 8 to Fig. 14 uses a reactor containment vessel called RCCV, but the type of reactor containment vessel is not limited to RCCV. It is universally applicable to all pressure suppression type reactor containment vessels that have a pressure suppression function using a pressure suppression pool. In addition, other materials such as SC structure and steel can also be used.

[0151] Further, the embodiment shown in Figs. 8 to 14 includes an external well 32. The material of the external well 32 can be any of the same materials as the reactor containment vessel 3, such as reinforced concrete (RC), SC, and steel. If the external well 32 is made of reinforced concrete, a liner is laid on the inner surface, as in the case of the reactor containment vessel 3. In the embodiment shown in Figs. 8 to 14, the planar shape of the external well 32 is rectangular and configured to surround part of the outer walls of the dry well 4 and the wet well 5, but the planar shape of the external well 32 is not limited thereto. Any shape may be used as long as the external well 32 is adjacent to and surrounds at least part of the outer walls of the dry well 4 and the wet well 5. For example, a trapezoid, a polygon, a crescent shape, a partial ring shape, a complete ring shape, and the like may be used. EXAMPLES

[0152] Fig. 8 is an elevational sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to the eighth embodiment of the present invention. The eighth embodiment differs from the conventional SATOR (see Fig. 21) in that, like the first embodiment, a reactor pressure vessel water injection pipe 51, a reactor pressure vessel water injection pipe isolation valve 51a, and a condensate return pipe isolation valve 21a are added. This allows the passive containment vessel cooling system 12 to be used in the reactor pressure vessel water injection mode in the event of a severe accident.

[0153] Compared to ABWR, SATOR has wetwell gas supply piping 48, outer well 32, and scrubbing pool 33, and is a nuclear plant that has achieved remarkable progress in reducing the pressure of the reactor containment vessel 3 and suppressing the leakage of radioactive materials in the event of a severe accident. Even if high-pressure steam, hydrogen, and radioactive materials are generated in the wetwell gas phase 7, the steam can be condensed by the passive containment vessel cooling system 12, and particulate radioactive materials can be removed by the scrubbing pool 33, and hydrogen, radioactive rare gases, and organic iodine can be trapped in the outer well 32. This makes it possible to suppress the pressure of the reactor containment vessel 3 in the event of a severe accident to below the design pressure, eliminating the possibility of overpressure damage to the reactor containment vessel 3, and making emergency evacuation of surrounding residents unnecessary. In addition, since most of the piping of the passive containment vessel cooling system 12 is installed inside the outer well 32, even if radioactive materials leak from the system, they can be retained inside the outer well 32. In addition, since there is always a pressure difference between the wetwell gas phase 7 and the outer well 32, a gas vent fan is not required. The eighth embodiment makes it possible to more reliably prevent overheating and damage in the event of a severe accident. EXAMPLES

[0154] FIG. 9 is an elevational cross-sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to the ninth embodiment of the present invention. The ninth embodiment differs from the conventional SATOR in that a backflow prevention device (water seal tank) 52 and a wetwell spray sparger 53 are further added, as in the second embodiment. This allows the passive containment cooling system 12 to be used in wetwell spray mode in the event of a severe accident. This makes it possible to prevent temperature stratification of the pressure suppression pool in the event of a severe accident. The other configurations of the passive containment cooling system 12 of the ninth embodiment are the same as those of the eighth embodiment. EXAMPLES

[0155] Fig. 10 is an elevational sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to the tenth embodiment of the present invention. The tenth embodiment differs from the conventional SATOR in that, like the third embodiment, the tip of the reactor pressure vessel water injection pipe 51 is further connected to the high pressure core cooling system sparger 47 in the shroud head 2b (see Fig. 19). This makes it possible to directly spray condensed water in the passive containment vessel cooling system 12 onto the debris in the reactor in the event of a severe accident. The other configurations of the passive containment vessel cooling system 12 in the tenth embodiment are the same as those in the ninth embodiment. EXAMPLES

[0156] 11 is an elevational sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to an eleventh embodiment of the present invention. The eleventh embodiment differs from the conventional SATOR in that, like the fourth embodiment, the reactor pressure vessel water injection pipe 51 is further connected to the high pressure core cooling system water injection pipe 95 and connected to the high pressure core injection system sparger 47 through a shared pipe. This eliminates the need to provide a nozzle for the reactor pressure vessel water injection pipe 51 in the reactor pressure vessel 2. The other configurations of the passive containment vessel cooling system 12 of the eleventh embodiment are the same as those of the tenth embodiment. EXAMPLES

[0157] FIG. 12 is an elevational cross-sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to the twelfth embodiment of the present invention. The twelfth embodiment differs from the conventional SATOR in that, like the fifth embodiment, a reactor pressure vessel gas supply pipe 54, a reactor pressure vessel gas supply pipe isolation valve 54a, and a pressure equalization valve 55 are further added. In the case of SATOR, the water depth of the pressure suppression pool is deep, so the pressure equalization valve 55 is not essential. This makes it possible to use the passive containment vessel cooling system 12 in the reactor pressure vessel steam cooling mode in the event of a severe accident. The air-cooling effect of the steam makes it possible to sufficiently cool the remaining debris in the reactor. The other configurations of the passive containment vessel cooling system 12 of the twelfth embodiment are the same as those of the eleventh embodiment. EXAMPLES

[0158] FIG. 13 is an elevational cross-sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to the thirteenth embodiment of the present invention. The thirteenth embodiment differs from the conventional SATOR in that, as in the sixth embodiment, a drywell spray piping 56, a drywell spray piping isolation valve 56a, and a drywell spray sparger 56b are further added. The drywell spray sparger 56b can use a sparger for the reactor containment vessel spray for the RHR (not shown). This allows the passive containment vessel cooling system 12 to be used in the drywell spray mode in the event of a core meltdown accident after a main steam pipe rupture, and allows the decay heat of radioactive materials attached to the wall and floor of the upper drywell 4c to be cooled and the temperature of the upper drywell 4c to be sufficiently lowered. The other configurations of the passive containment vessel cooling system 12 of the thirteenth embodiment are the same as those of the twelfth embodiment. EXAMPLES

[0159] FIG. 14 is an elevational cross-sectional view showing the configuration around the reactor containment vessel of a nuclear power plant according to the 14th embodiment of the present invention. The 14th embodiment differs from the conventional SATOR in that, as in the 7th embodiment, a backflow prevention device (water seal tank) 52 is further provided upstream of the connection point between the drywell spray piping 56 and the condensate return piping 21. This makes it possible to prevent gas in the drywell 4 from flowing back from the drywell spray piping 56 into the heat exchanger 16 of the passive containment cooling system 12. This makes it possible to operate the passive containment cooling system 12 in the drywell spray mode in the event of a loss of coolant accident by opening the drywell spray piping isolation valve 56a during normal operation, and to depressurize the drywell 4 immediately after the accident occurs. The other configurations of the passive containment cooling system 12 of the 14th embodiment are the same as those of the 13th embodiment.

[0160] (Other embodiments) Although some embodiments of the present invention have been described, these embodiments are presented as examples and are not intended to limit the scope of the invention. These embodiments can be implemented in various other forms, and various omissions, substitutions, and modifications can be made without departing from the spirit of the invention. These embodiments and their modifications are included in the scope of the invention and its equivalents described in the claims, as well as in the scope and spirit of the invention.

[0161] (Figure 26: Description of features common to the embodiments) Hereinafter, the features of the configuration common to the passive containment cooling system 12 of the first to fourteenth embodiments will be described with reference to FIG. 26. In the embodiment of the present invention, at least one gas supply pipe 29 is provided, one end of which is connected to the inlet plenum 17 of the heat exchanger 16 and the other end of which is connected to the inside of the reactor containment vessel 3, and which guides the gas inside the reactor containment vessel 3 to the heat exchanger 16. In FIG. 26, one end of the gas supply pipe 29 is connected to the inside of the upper dry well 4c, but this example is not limited thereto. The gas inside the reactor containment vessel 3 includes the gas inside the reactor pressure vessel 2 and the gas in the wet well gas phase 7. In addition, a condensed water return pipe 21 is provided, one end of which is connected to the outlet plenum 18 of the heat exchanger 16 and the other end of which is connected to the lower region 90 inside the reactor containment vessel 3, and which guides the condensed water inside the heat exchanger 16 into the lower region 90 inside the reactor containment vessel 3. In FIG. 26, the condensed water return pipe 21 is led to the pressure suppression pool 6, but this example is not limited thereto. The lower region 90 includes the wetwell gas phase section 7 and the lower drywell 61a. Furthermore, in the embodiment of the present invention, at least one upper region water injection pipe 91a is provided that branches from the condensate return pipe 21 and injects condensate into the upper region 91 of the reactor containment vessel 3. In FIG. 26, the upper region water injection pipe 91a is connected to the upper drywell 4c, but this example is not limited to this. The upper region 91 includes equipment in the upper region 91. For example, the upper region 91 includes the reactor pressure vessel 2. Furthermore, a gas vent pipe 22 is provided, one end of which is connected to the outlet plenum 18 of the heat exchanger 16 and which discharges gas in the outlet plenum to the outside of the heat exchanger 16. Here, the pressure suppression pool 6 and the scrubbing tank 33 are included outside the heat exchanger 16. Note that in FIGS. 1 to 14, the configurations of the reactor core 1, the reactor pressure vessel 2, and the reactor containment vessel 3 are the same as those of the ABWR, but are not limited to this. The passive containment cooling system 12 according to the present invention may be used in any nuclear power plant that utilizes a pressure suppression type reactor containment. [Explanation of symbols]

[0162] 1...core, 2...reactor pressure vessel (RPV), 2a...shroud, 2b...shroud head, 2c...reactor pressure vessel head, 2d...downcomer, 2e...steam dome, 2f...lower plenum, 2g...reactor pressure vessel bottom head (lower head), 2s...steam phase, 3...reactor containment vessel, 4...drywell, 4a...top slab, 4b...drywell common wall, 4c...upper drywell, 5...wetwell, 5a...wetwell common wall, 5b...diaphragm floor, 6...pressure suppression pool, 7...wetwell gas phase, 8...LOCA vent pipe, 8a...horizontal vent pipe, 9 ...Vacuum breaker valve, 10...Primary containment vessel top head (PCV top head), 10a...Primary containment vessel top head (PCV top head) part, 11...Water shield, 12...Passive containment vessel cooling system (PCCS), 13...Cooling water pool, 14...Cooling water, 15...Exhaust port, 16...Heat exchanger, 17...Inlet plenum, 18...Outlet plenum, 19...Heat transfer tube, 20...Drywell gas supply piping, 20a...Drywell gas supply piping isolation valve, 21...Condensate return piping, 21a...Condensate return piping isolation valve, 22...Gas vent piping, 22a...Gas vent fan, 22b...Backflow prevention valve, 23...Tube sheet, 24...PCCS drain tank, 25...debris protection cover, 26...RPV water injection piping, 27...RPV water injection isolation valve, 28...GDCS pool, 29...gas supply piping, 32...outer well, 32a...top slab, 33...scrubbing pool, 33a...top cover, 33b...space, 33c...first outlet piping, 34...metal fiber filter, 34a...second outlet piping, 40...internal pump, 41...pump deck, 42...diffuser, 43...steam separator, 44...steam dryer, 45...steam dryer skirt, 46...low pressure water injection system sparger, 47...high pressure Core injection system sparger, 48...wetwell gas supply piping, 50...water supply sparger, 51...reactor pressure vessel (RPV) injection piping, 51a...reactor pressure vessel (RPV) injection piping isolation valve, 52...backflow prevention device (water tank), 53...wetwell sparger, 54...reactor pressure vessel (RPV) gas supply piping, 54a...reactor pressure vessel (RPV) gas supply piping isolation valve, 55...pressure equalization valve, 56...drywell spray piping, 56a...drywell spray piping isolation valve, 56b...drywell spray sparger, 61...pedestal,61a...pedestal cavity (lower dry well), 62...RPV skirt (vessel skirt), 63...RPV support (vessel support), 64...melting valve, 65...lower dry well flooding piping, 66...opening, 67...flooding valve, 68...flooding piping, 69...check valve, 71...main steam piping, 72...safety relief valve, 73...relief piping, 74...flange, 75...platform , 76...manhole, 77...ventilation duct, 90...lower region, 91...upper region, 91a...upper region water injection piping, 91b...upper region water injection piping isolation valve, 92...high pressure core injection system (HPCF), 93...high pressure core injection system pump, 94...high pressure core injection system suction piping, 95...high pressure core injection system water injection piping, 96...high pressure core injection system outer isolation valve, 97...high pressure core injection system inner isolation valve, 100...reactor building,

Claims

1. The core, a reactor pressure vessel containing the reactor core; a shroud surrounding the reactor core within the reactor pressure vessel; a shroud head provided on an upper portion of the shroud; a reactor containment vessel including a drywell that houses the reactor pressure vessel, a wetwell that houses a pressure suppression pool in a lower portion and is connected to the drywell via a LOCA vent pipe and has a wetwell gas phase in an upper portion, and a vacuum breaker valve that circulates gas in the wetwell gas phase to the drywell; a pedestal that supports the reactor pressure vessel within the reactor containment vessel via an RPV skirt and an RPV support and defines a pedestal cavity therein; an upper drywell which is a space above the RPV skirt of the drywell; A top slab constituting the ceiling of the upper drywell; a diaphragm floor constituting the ceiling of the wet well and constituting the floor of the upper dry well; an upper region of the reactor containment vessel above the diaphragm floor of the upper drywell and below the top slab; a lower region of the reactor containment vessel below the diaphragm floor; a cooling water pool installed on an upper portion of the dry well to store cooling water; a heat exchanger having an inlet plenum, an outlet plenum, and heat transfer tubes, at least a portion of which is submerged in the cooling water; at least one gas supply pipe having one end connected to the inlet plenum of the heat exchanger and another end connected to the inside of the reactor containment vessel for guiding gas inside the reactor containment vessel to the heat exchanger; a gas supply pipe isolation valve provided in the gas supply pipe; a condensate return pipe having one end connected to the outlet plenum of the heat exchanger and another end connected to a lower region within the reactor containment vessel for directing condensate within the heat exchanger to the lower region within the reactor containment vessel; a condensate return pipe isolation valve provided in the condensate return pipe; At least one upper region water injection pipe branching from the condensate return pipe for injecting condensate into an upper region of the reactor containment vessel; an upper region water injection piping isolation valve provided in the upper region water injection piping; a gas vent pipe having one end connected to the outlet plenum of the heat exchanger and configured to discharge gas in the outlet plenum to the outside of the heat exchanger; a passive containment cooling system having a A nuclear power plant comprising:

2. 2. The nuclear power plant according to claim 1, further comprising a passive containment vessel cooling system, wherein one end of the upper region water injection piping is a reactor pressure vessel water injection piping connected to a reactor pressure vessel for injecting condensed water into the reactor pressure vessel.

3. 2. The nuclear power plant according to claim 1, further comprising a passive containment vessel cooling system, wherein one end of said gas vent piping is immersed in and connected to said pressure suppression pool, and a gas vent fan is provided on said gas vent piping.

4. 2. The nuclear power plant according to claim 1, having a passive containment vessel cooling system, characterized in that a backflow prevention device is provided in the middle of the condensate return pipe, an end of the condensate return pipe is connected to the wetwell gas phase portion, and the condensate return pipe can spray condensed water in the heat exchanger to the wetwell gas phase portion.

5. 2. The nuclear power plant according to claim 1, further comprising a reactor pressure vessel water injection pipe, one end of which is connected to a shroud head in the reactor pressure vessel and capable of injecting condensed water into the shroud head.

6. 2. The nuclear power plant according to claim 1, further comprising a passive containment vessel cooling system, wherein one end of the upper region water injection piping is connected to a high pressure core water injection system water injection piping of a high pressure core water injection system.

7. 2. The nuclear power plant according to claim 1, further comprising a passive containment vessel cooling system, wherein one end of the gas supply pipe is a reactor pressure vessel gas supply pipe connected to the reactor pressure vessel and supplying gas in the reactor pressure vessel to the heat exchanger.

8. 2. The nuclear power plant according to claim 1, further comprising a pressure equalizing valve connected to the steam phase of the reactor pressure vessel via a pipe provided in the drywell, so that the pressure in the reactor pressure vessel can be equalized with the pressure in the drywell when the reactor pressure vessel is damaged and the pressure in the reactor pressure vessel drops.

9. 2. The nuclear power plant having a passive containment vessel cooling system according to claim 1, wherein the upper region water injection piping is a drywell spray piping connected to a drywell spray sparger in the upper drywell for spraying condensed water into the upper drywell.

10. 9. The nuclear power plant according to claim 8, having a passive containment vessel cooling system, characterized in that a backflow prevention device is provided in the condensate return piping upstream of a branch point of the drywell spray piping, thereby preventing gas in the drywell from flowing back from the drywell spray piping and enabling the drywell spray piping isolation valve to be opened during normal operation.

11. an outer well provided outside the dry well and the wet well and adjacent to the dry well via a dry well common wall and adjacent to the wet well via a wet well common wall, completely surrounding the dry well common wall and the wet well common wall and having pressure resistance and airtightness equivalent to those of the dry well and the wet well; a scrubbing pool provided in the external well and storing water therein; a cooling water pool installed on the upper portion of the dry well and the outer well to store cooling water; a heat exchanger having an inlet plenum, an outlet plenum, and heat transfer tubes, at least a portion of which is submerged in the cooling water; a wetwell gas supply pipe having one end connected to the inlet plenum of the heat exchanger, passing through the external well, and the other end connected to the wetwell gas phase portion for directing gas in the wetwell gas phase portion to the heat exchanger; a gas vent pipe having one end connected to the outlet plenum of the heat exchanger, passing through the external well, and the other end immersed in the water in the scrubbing pool in the external well, for discharging gas in the heat exchanger to the external well; 11. The nuclear power plant according to claim 1, further comprising a passive containment cooling system, wherein the gas supply piping, the condensate return piping and the upper zone water injection piping at least partially pass through the interior of the outer well.

Images