JP2000292580A - Nuclear power plant - Google Patents

Abstract

(57) [Summary] [PROBLEMS] To provide a compact nuclear power generation facility capable of simultaneously improving economic efficiency and safety and significantly improving competitiveness with other alternative power generation facilities of nuclear power generation. A nuclear power plant has a structure in which a reactor vessel is housed in a reactor vessel, and a main steam pipe connected to the reactor vessel is led out through the reactor vessel. The plurality of main steam isolation valves 21 and 22 provided in the main steam pipe 20 are all installed outside the reactor containment vessel 8.

Description

DETAILED DESCRIPTION OF THE INVENTION

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a light water nuclear power plant (LWR), and more particularly to a nuclear power plant with a great improvement in economy and safety.

2. Description of the Related Art A typical example of a conventional LWR is a boiling water reactor (BWR). Below,
The prior art will be described by taking the case of BWR as an example. Conventional BWRs are all contained in a reactor containment vessel (PCV).
Contains one reactor pressure vessel (RPV). Among the latest BWRs, the improved BWR (ABW
R) are known. The reactor containment vessel of this ABWR is made of reinforced concrete called so-called RCCV, and is the most sophisticated one currently in practical use. FIG. 11 schematically shows the installation status of the reactor containment vessel and the like of the ABWR. As shown in FIG. 11, the reactor pressure vessel 1 is installed in a dry well (DW) 2, and the dry well 2 is further divided into an upper dry well (UD) 3 and a lower dry well (LD) 4. The dry well 2 is connected to a suppression pool (S / P) 7 in a wet well (WW) 6 by a vent pipe 5. The containment vessel 8 is thus mainly constituted by the dry well 2 and the wet well 6. As a design standard accident of the reactor containment vessel 8, it is assumed that a primary system pipe (main steam pipe) 9 connected to the reactor pressure vessel 1 breaks in the reactor containment vessel 8. This is generally referred to as a coolant loss accident (L
OCA). Even if a coolant loss accident occurs,
The high-temperature steam discharged into the dry well 2 immediately passes through the vent pipe 5 and is condensed in the suppression pool 7, so that the internal pressure of the containment vessel 8 is designed to be sufficiently low. In the conventional configuration, the main steam pipe 9
A first main steam isolation valve (first MSIV) 21 is installed inside the containment vessel 8. An emergency core cooling system (ECCS) 10 using a pump 10a, which is a dynamic device, is installed to replenish the coolant that is lost in the event of a coolant loss accident, and pumps pool water in the suppression pool 7. Is designed so that water can be injected into the reactor pressure vessel 1. In order to ensure reliability, a plurality of emergency core cooling systems, including those with different operating principles, are installed, and are designed to be extremely safe with multiplexing and diversity secured ( In the figure, only one system is shown for simplicity). Further, a residual heat removal system (RHR) 11 is installed for the purpose of removing decay heat generated continuously after the accident from the reactor core. This residual heat removal system uses a pump (RH) which is also a dynamic device.
The suppression pool water is sucked by using the R pump 11a, and is cooled and cooled by the heat exchanger (RHR heat exchanger) 11b. It is designed to provide cooling. The residual heat removal system 11 is a safety system having a safety function as described above, but is also used as a cooling system when the reactor is normally shut down. In order to operate a dynamic safety system such as an emergency core cooling system or a residual heat removal system, an emergency diesel generator (emergency D / G) 12 is used as an emergency power source.
It is installed (only one is shown in the figure).
The residual heat removal system 11 shares the pump with the emergency core cooling system and forms a part of the emergency core cooling system. On the other hand, in recent years, safety systems such as emergency core cooling systems and residual heat removal systems have all been performed only by static means, while dynamic devices such as pumps have been excluded from safety systems, although they have not been put into practical use in recent years. Static safety furnace (SBW)
R) has been proposed. Hereinafter, according to FIG.
Will be described. In FIG. 12, the same portions as those in FIG. 11 are denoted by the same reference numerals as those in FIG. 11, and the description of the overlapping portions will be omitted, and only the main portions will be described. In FIG. 12, a gravity falling emergency core cooling system (GDCS) 13 which is a static safety system is installed in the upper dry well 3. A static containment cooling system (PCCS) 14 and an isolation condenser (IC) 15 are provided above the reactor containment vessel 8. Assuming the occurrence of a coolant loss accident, which is a design standard accident, the point at which steam condenses in the suppression pool 7 is the above-mentioned ABW.
Similar to the case of R, the supply of the coolant into the reactor pressure vessel 1 is performed by gravity-dropping of the GDCS pool water 16 by gravity. Further, the cooling of the reactor containment vessel 8 is also performed by the steam and the like in the reactor containment vessel 8 being guided into the PCCS heat exchanger (PCCSHx) 17 by a slight differential pressure and cooled by the PCCS pool 18. Similarly, when the main steam isolation valve 21 is closed and the reactor is in an isolated state, cooling is performed by guiding the reactor steam in the reactor pressure vessel 1 to the IC heat exchanger (ICHx) 19 by buoyancy to condense. Will be Thus, SBWR
In the case of (1), when any abnormal situation occurs in the nuclear reactor, all the devices are statically performed using natural force. Therefore, these safety systems are simplified, reliability is greatly improved, and there is great expectation that an emergency power supply for operating them will not be required. Also in this example, the first main steam isolation valve (first M
SIV) 21 is provided. Further, in recent years, as a type of a powerful reactor containment vessel for the next-generation ABWR, a type having a suppression pool called a raised suppression pool has been studied (exhibition: Energy Review, January 1998, p. 19). An outline of the raised suppression pool will be described with reference to FIG. In FIG. 13, the same parts as those in FIGS. 11 and 12 are denoted by the same reference numerals as those in FIGS. 11 and 12, and the description of the overlapping parts is omitted, and only the main parts will be described. As shown in FIG. 13, the dry well 2 forms an integral space,
There is no distinction between upper and lower drywells. The main steam pipe (MS pipe) 20 hangs down in the dry well 2, and is guided from the lower part of the dry well 2 to the outside of the containment vessel 8 through a tunnel room (MS tunnel room) 23. The first main steam isolation valve 21 of the main steam pipe 20 is provided on the dry well 2 side, and the second main steam isolation valve 22 is arranged in a tunnel chamber 23. Further, the suppression pool 7 is located considerably above the floor of the containment vessel 8,
Safety equipment such as an emergency core cooling system can be housed in a lower chamber (ECCS chamber) 24. Wet well 6
Extends to the upper end of the containment vessel 8 and is at the same height as the top slab (top plate) of the dry well 2. In the figure, reference numeral 25 denotes a reactor containment vessel boundary (P
CV boundary). In the raised suppression pool configured as described above, the vapor phase space of the wet well 6 is large, and it is expected to improve the safety in that the pressure rise at the time of an accident can be suppressed lower. Further, since safety equipment such as an emergency core cooling system can be housed in the ECCS room 24 in the lower part of the containment vessel 8, the floor area of the reactor building can be reduced, and a great improvement in arrangement can be expected.

However, the above-described technique has various problems to be solved. That is, FIG.
In the case of the safety system of the static safety reactor (SBWR) shown in
It requires a large amount of pool water for cooling, which has a significant effect on layout. For example, PCCS pool 18 is about 250
0 m ³ . For this reason, the reactor building cannot be downsized, which is an economical disadvantage. Furthermore, the gravity drop type emergency core cooling system (GDCS) 13 has only a water injection capability with a head having a slight difference in elevation, and it is necessary to rapidly depressurize the reactor pressure vessel 1 in the event of a coolant loss accident. For this reason, it is necessary to install a large number of pressure reducing valves. Considering the failure of operation of the pressure reducing valve, a gravity drop type emergency core cooling system (GD
It is difficult to make the reliability of the CS) 13 itself as an emergency core cooling system extremely high. In addition, although the safety system is all static, the residual heat removal system for normal shutdown needs to leave the conventional dynamic device. Residual heat removal systems, including reactor auxiliary cooling system (RCW) and reactor auxiliary equipment seawater cooling system (RSW) (hereinafter referred to as "support system"), are expensive in safety systems, and eventually However, it has not led to the rationalization of safety systems. Furthermore, although an emergency power supply would have been unnecessary, in practice, some safety equipment had to be configured with dynamic devices, so some emergency power supply had to be left. As described above, in SBWR, there are issues such as the influence of the arrangement by the static safety system and the further rationalization of the dynamic system. To mitigate placement effects,
Reducing the size of the PCV is also an important issue. Note that S
The BWR is not yet in operation at present. On the other hand, A
The BWR has already been put into practical use and has been recognized for its excellent safety. However, due to discussions on responding to severe accidents in recent years, some static It has been studied to add a static safety system such as a containment cooling system to further enhance the safety. In the design of such a next-generation reactor, the influence on the layout of the reactor building becomes significant. To alleviate this, it is important to further reduce the volume of the containment vessel. Conversely, as the size of the containment vessel and reactor building of the ABWR has been reduced, the minimum required area of the operating floor of the reactor building has been secured, resulting in restrictions on operation during regular inspections. ing. Accordingly, it is important to increase the area of the operating floor of the reactor building while alleviating the influence of the layout of the entire building. Although not shown, the present applicant has proposed a lower drywell separation method as a modified reactor containment vessel of the ABWR (for example, Japanese Patent Application No. 10-87411). According to this lower drywell separation method, the wetwell gas phase volume can be significantly reduced, but still the upper drywell cannot be made too small to secure the installation space for the main steam isolation valve. There was a limit to the volume reduction of the entire PCV. For this reason, there is a situation to be studied until a plan to reduce the volume of the reactor containment vessel while securing the installation space for the main steam isolation valve by eccentricizing the reactor containment vessel. That is, even in the case of the lower drywell separation method, it is important to reduce the inner diameter of the upper drywell. Further, in the case of the raised suppression pool shown in FIG. 13, which has been studied as a possible type of a primary containment vessel for the next-generation ABWR, the main steam pipe 20 passes through the lower part of the dry well 2 so that The space under the well 2 becomes considerably larger than the conventional containment vessel. Since the suppression pool 7 is located at the upper part, when the emergency core cooling system is activated at the time of an accident, the suppression pool water stays in the lower space of the vast dry well 2 and does not return to the suppression pool 7. Although this is referred to as drawdown water, this to ensure from the beginning of the loss amount of the water source, about 7000 m ³ of the suppression pool water from traditional about 3500 m ³ of approximately twice
Need to be increased to As a result of considering the result,
The wet well volume has increased, and eventually the total volume of the containment vessel has increased, leading to a decrease in economic efficiency.
Therefore, in the case of a raised suppression pool, it is necessary to minimize the volume of the space under the dry well 2 as much as possible.
It becomes impossible to secure a sufficient space for installing or maintaining a large device such as the above. Originally, the concept of a reactor containment vessel that was originally developed with an emphasis on the improvement of the placeability could result in a loss of its feasibility due to serious problems in the place. Therefore, in the case of a raised suppression pool, it is very important to make the space under the drywell more compact without sacrificing the arrangement of such main devices. The present invention has been made in view of such circumstances, and provides a compact nuclear power plant capable of simultaneously improving economic efficiency and safety, and further significantly improving the competitiveness with other alternative power plants. The purpose is to provide.

According to one aspect of the present invention, a reactor vessel is housed in a reactor containment vessel, and a main steam pipe connected to the reactor vessel is provided. In the nuclear power plant configured to penetrate through the reactor containment and lead out to the outside, a plurality of main steam isolation valves provided in the main steam pipe are all installed outside the reactor containment. And That is, regarding the two main steam isolation valves, the first main steam isolation valve and the second main steam isolation valve, which have been installed at different locations in the reactor containment vessel and the exterior of the reactor containment vessel, have been separated. In the invention, both valves are installed outside the containment vessel. Therefore, according to the first aspect of the present invention, since the main steam isolation valve is released from extremely severe environmental conditions such as temperature, pressure, and radioactivity that are exposed in a severe accident, for example, Can be increased. Also, conventionally, the main steam isolation valve provided in the reactor containment vessel was the largest factor determining the drywell inner diameter except for the reactor pressure vessel body, but by removing this, The inner diameter of the dry well can be greatly reduced. As the inner diameter of the drywell becomes smaller, the volume of the drywell becomes smaller.This not only greatly reduces the volume of the reactor but also reduces the pressure in the reactor containment vessel in the event of a loss of coolant. Has a positive effect. That is, when the dry well volume is reduced, the amount of non-condensable gas such as nitrogen existing before the accident can be reduced, and in the event of a coolant loss accident, nitrogen or the like moves to the wet well gas phase and is compressed. This is because the rise in the internal pressure of the containment vessel due to the above can be further reduced. The invention according to claim 2 is directed to a raised suppression pool type reactor containment vessel that leads a main steam pipe to the outside of the reactor containment vessel through a suppression pool located above the lowermost part of the containment vessel. In the nuclear power plant having, a plurality of main steam isolation valves provided in the main steam pipe are all installed outside the containment vessel,
The drawdown space below the reactor containment vessel is minimized. According to the present invention, two main steam isolation valves are installed in the tunnel room on the outer peripheral side of the dry well, and when applied to a raised suppression pool, the space below the dry well is reduced in the arrangement of the main steam isolation valve. It can be made sufficiently small without sacrificing.
As a result, the drawdown water amount is greatly reduced, and the suppression pool water amount is also reduced, so that the wet well volume is reduced, the reactor containment volume is reduced, and the cost of the reactor containment can be reduced. The invention according to claim 3 is the nuclear power plant according to claim 1 or 2, wherein, in the main steam pipe, a pipe portion from a penetrating portion of the containment vessel to a main steam isolation valve disposed outside the main containment vessel, It has a double structure. In the present invention, when the two main steam isolation valves are installed outside the reactor containment vessel or in the tunnel room, the main steam pipe between the second main steam isolation valve and the containment vessel has a double piping structure. So, temporarily
Even if the main steam pipe up to the main steam isolation valve is broken, it has a double structure, so that it is possible to eliminate the influence of a pipe break accident. As a result, sufficient safety can be ensured even when the two main steam isolation valves are installed outside the reactor containment vessel. According to a fourth aspect of the present invention, in the boiling water nuclear power plant, the control rod drive mechanism connected to a lower part of the reactor pressure vessel is drawn upward in the reactor pressure vessel, whereby the control rod The lower space of the drive mechanism is deleted, and the reactor pressure vessel is made to approach the lowermost part of the containment vessel. That is, in the present invention, the control rod drive mechanism (FMCRD) which is installed hanging down at the lower part of the reactor pressure vessel, which has been pulled out at the time of conventional maintenance or the like, is changed to the inside of the reactor pressure vessel. In other words, by pulling out to the upper side, most of the lower drywell space of, for example, about 11 m conventionally installed in the lower part of the reactor pressure vessel is deleted, and the reactor pressure vessel is brought as close as possible to the lower floor of the containment vessel. And install it. Therefore, according to the present invention, since the position of the reactor pressure vessel, which is the heaviest object, shifts significantly downward, the seismic conditions of the core fuel are greatly relaxed, and the seismic resistance of the reactor plant is greatly improved. be able to. Furthermore, while the height of the reactor containment vessel was conventionally determined as the sum of the reactor pressure vessel height and the lower drywell height, according to the present invention, the lower drywell height is reduced by half. Thereby, the height of the reactor containment vessel can be significantly reduced. As a result, the reactor containment vessel can be rationalized, and further, the reactor building itself can be rationalized. By reducing the height of the reactor building, for example, the reactor building can be reduced by one level, whereby the plant construction process can be greatly shortened. For example, it is possible to reduce the world's shortest 48-month process to the level of a 36-month process even today. The invention according to claim 5 is a boiling water nuclear power plant, wherein two or more reactors are provided in the same facility, and a safety system including one wet well and a suppression pool is shared by the two or more reactors. It is characterized in that it is configured to be. According to the present invention, in a reactor containment vessel of a boiling water nuclear power plant, since a wet well and a suppression pool are shared by two or more reactors, a wet well conventionally provided for each reactor is provided. And a single suppression pool, greatly reducing the overall volume of the containment vessel,
The arrangement inside the reactor building can be greatly reduced. According to a sixth aspect of the present invention, in the nuclear power plant according to the fifth aspect, a midwell for suppressing pressure is newly provided above the wet well, and the midwell and a suppression pool in the wetwell are vented. Are connected. According to the present invention, a shared space called a midwell is newly provided, and this and two or more dry wells are connected by a vent pipe, and the midwell is connected to a suppression pool in a wet well by a vent pipe. This eliminates the need to separately connect vent tubes from two or more dry wells to wet wells.
Thereby, only the vent pipe for one plant needs to be connected from the midwell to the wet well. That is, if the vent pipes for the two plants are connected directly to the wet wells independently of each other, the number of vent pipes guided into the suppression pool becomes considerably large, and the wet wells need to be enlarged. In the present invention, this can be avoided. According to a seventh aspect of the present invention, in the nuclear power plant according to the fifth or sixth aspect, a dry well in each nuclear reactor and one shared wet well or midwell are connected by a connecting vent pipe, The connecting vent pipe is provided with a check valve or a rupture disk having a pressure support mechanism on the back side. According to the present invention, a check valve or a rupture disk provided with a pressure support mechanism on the back surface is provided in the vent pipe so that the dry well and the wet well do not come into direct contact with each other in normal times. Even if the dry well is opened during the regular inspection, it is possible to prevent any safety problems from occurring. Since the rupture disk is supported on the back side by the pressure indicating mechanism, there is no possibility that the rupture disk will malfunction even when a back pressure is applied. In addition, if a coolant loss accident occurs in one of the reactors, the check valve or rupture disk of the connecting vent pipe installed between the reactor and the dry well of the healthy side may malfunction and the impact of the accident may be reduced. It is possible to avoid reaching the healthy reactor side. The invention according to claim 8 is the light water nuclear power plant, wherein two or more reactors are provided in the same facility, and one static safety system is shared by the two or more reactors. Features. Here, the static safety system includes a static containment vessel cooling system, an isolation condenser, a gravity drop type emergency core cooling system, and the like. ADVANTAGE OF THE INVENTION According to this invention, it becomes possible to remarkably reduce the influence of arrangement | positioning of the static safety system which had the great influence conventionally. The present invention can be applied to all light water nuclear power plants using a static safety system. According to a ninth aspect of the present invention, in a light water nuclear power plant, two or more nuclear reactors are provided in the same facility,
The pit is shared by the two or more nuclear reactors. Here, the dryer separator pit (DS pit) refers to a storage pit such as a steam dryer and a steam separator. According to the present invention, it is not necessary to install a dryer / separator pit for each plant,
The operating floor can be used more widely. Further, it is not necessary to unnecessarily increase the floor area of the reactor building. Conventionally, the fuel pool and the dryer / separator pit are arranged in a straight line, and if these are arranged in series for two plants, the size of the reactor building becomes very long. According to the present invention, this is avoided. It becomes possible. That is, since the dryer / separator pit is used at the time of refueling of the plant, it is not necessary to install the dryer / separator pit for each plant unless simultaneous inspection of the two plants is performed. In addition, it is possible to respond to the requirement for simultaneous inspection of two plants by securing the capacity of the dryer / separator pit so that only the steam-water separator that needs to be stored underwater can store two plants. It is. The invention according to claim 10 is a nuclear power plant equipped with a high-pressure core cooling system, a low-pressure core cooling system, a residual heat removal system, and a reactor accessory cooling system as an emergency core cooling system, wherein the high-pressure core cooling system includes: The low-pressure core cooling system and the residual heat removal system are divided into four sections, while the piping of the reactor auxiliary equipment cooling system is divided into two sections.
It is characterized by a loop configuration. According to an eleventh aspect of the present invention, in the nuclear power plant according to the tenth aspect, two or more reactor auxiliary unit cooling system pumps and a reactor auxiliary unit seawater cooling system are provided in piping of each loop of the reactor auxiliary unit cooling system. , And the emergency power supply for supplying power to these is divided into four sections. These claims 10, 1
According to the first aspect of the invention, all the front lines of the emergency core cooling system including the high-pressure core cooling system are divided into four sections, so that all of the front lines and the support lines of the emergency core cooling system can be maintained online. Becomes The invention according to claim 12 is characterized in that the emergency core cooling system according to claim 10 or 11 is configured to be shared by two or more nuclear reactors. According to the present invention, all parts of the emergency core cooling system can be maintained on-line, so that maintenance of the emergency core cooling system can be performed at any time without performing two plants at the same time. According to a thirteenth aspect of the present invention, both the configuration of the main steam pipe according to any one of the first to third aspects and the configuration of the reactor pressure vessel and the containment vessel according to the fourth aspect are provided. It is characterized by. According to a fourteenth aspect of the present invention, there is provided a nuclear reactor having a configuration according to the thirteenth aspect. It is characterized by a combination of configurations.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of a nuclear power plant according to the present invention will be described below with reference to FIGS. First Embodiment FIG. 1 shows a configuration of a reactor containment vessel of a nuclear power plant and its periphery according to a first embodiment of the present invention.
In this embodiment, as shown in FIG. 1, a reactor pressure vessel 1 is installed in a dry well 2, and the dry well 2 is further separated into an upper dry well 3 and a lower dry well 4. The dry well 2 is connected to a suppression pool 7 in the wet well 6 by a vent pipe 5. The reactor containment vessel 8 is thus of a lower drywell separation type. In such a configuration, in the present embodiment, the main steam pipe 9 connected to the reactor pressure vessel 1 is used.
A plurality of main steam isolation valves provided in the main steam pipe 9 through the reactor containment vessel 8 to the outside.
For example, the first and second two main steam isolation valves 21 and 22
All are arranged outside the containment vessel 8. According to such a configuration, the first main steam isolation valve 21 conventionally disposed in the upper dry well 3 is disposed outside the reactor containment vessel 8 so that the main steam isolation valve 21 is disposed in the upper dry well 3. The space for the valve 21 is not required, so that the inner diameter of the upper dry well 8 can be greatly reduced. For example, the inner diameter of the dry well can be reduced from the current 29 m to about 20 m as shown in FIG. In addition, when the dry well volume is reduced by such a configuration, the amount of nitrogen, which is a non-condensable gas, existing in the dry well is reduced, and the amount of nitrogen gas contained in the wet well side in the event of a coolant loss accident is reduced. It will be greatly reduced. As a result, the effect of improving safety that the peak pressure of the containment vessel 8 at the time of the coolant loss accident can be reduced can be obtained. Conventionally, the first main steam isolation valve inside the containment vessel was exposed to severe environmental conditions in the containment vessel during a severe accident or the like. By installing the isolation valve 21 outside the containment vessel 8, this problem can be fundamentally solved. Further, the accessibility to the main steam isolation valve 21 is improved, and the dose of the operator exposed in the containment vessel 8 during maintenance can be significantly reduced. As shown in FIG. 1, in the present embodiment, the inner diameter of the upper dry well 3 is about 20 m and the inner diameter of the wet well 6 is about 26 m. This is for securing the pool water volume of the pool 7. As a modification of the present embodiment, for example, when the above-described main steam isolation valve arrangement is applied to a reactor containment vessel of a type in which a wet well is separately provided, the inner diameter of the reactor containment vessel is reduced. It is possible to be unitary as a whole and to match the inner diameter of the upper drywell. In that case, the shape of the containment vessel can be made a uniform cylindrical shape. Note that the dimensional values in FIG. 1 are merely examples, and can be variously changed. In addition, the main steam isolation valve 2 provided outside the reactor containment vessel 8 can be changed.
The number of 1, 2 can also be set to an arbitrary number of 2 or more. As described above, the dimensions of each part of the reactor containment vessel 8 and the number of valves and other parts are not necessarily limited to those shown in the drawings and the like, and various changes can be made in the following embodiments. is there. Second Embodiment FIG. 2 shows a reactor containment vessel of a nuclear power plant and its peripheral configuration according to a second embodiment of the present invention. In the present embodiment, as in the first embodiment, a plurality of, for example, first and second two main steam isolation valves 21 and 22 provided in the main steam pipe 9 are all provided outside the reactor containment vessel 8. Are located. Further, in the present embodiment, of the main steam pipe 9, a portion from the penetrating portion of the reactor containment vessel 8 to the first main steam isolation valve 21 installed outside the reactor containment vessel 8, The portion from the main steam isolation valve 21 to the second main steam isolation valve 22 has a double pipe structure covered with a guard pipe 8a. Other configurations are the same as those of the first embodiment, and therefore, the same reference numerals as in FIG. 1 are assigned to FIG. 2 and the description is omitted. According to the configuration of the present embodiment, the following effects are exerted in addition to the effects of the first embodiment. That is, conventionally, since the main steam pipe between the main steam isolation valves is the boundary of the reactor containment vessel, the assumption of the breakage of the pipe was not required in the safety design. However, considering the piping breakage at this portion, even if the second main steam isolation valve outside the reactor containment vessel is closed as designed, its isolation function is lost. Therefore, even in this case, the first main steam isolation valve has been installed inside the reactor containment vessel so that the containment vessel can be isolated. In contrast, in the present embodiment, the second steam pipe 8
Since the portion up to the main steam isolation valve 22 is protected by the double pipe structure by the guard pipe 8a, even if the piping is broken at this portion, the first main steam isolation valve 21 is connected to the reactor containment vessel 8 It is possible to eliminate the need for installation inside. Third Embodiment FIG. 3 shows a configuration of a containment vessel of a nuclear power plant according to a third embodiment of the present invention. This embodiment is applied to the above-mentioned raised suppression pool type reactor containment vessel, and the overall configuration is substantially the same as that shown in FIG. 13. Therefore, FIG. The description will be omitted, and only the main configuration will be described. As shown in FIG. 3, in the present embodiment, a main steam pipe 9 connected to the reactor pressure vessel 1 is led out of the containment vessel 8. In this embodiment, both the main steam isolation valves 21 and 22 are installed in the tunnel room 23 outside the reactor containment vessel 8 and
The pipe section on the upstream side of the main steam isolation valves 21 and 22 in 3 is covered with a guard pipe 9b to form a double pipe structure. According to such a configuration, the first main steam isolation valve 2
Are arranged outside the reactor containment vessel 8, so that the space below the dry well 2 is minimized by that much. For example, under such a configuration of the present embodiment, the inner diameter of the lower portion of the dry well 2 can be reduced from the current 17 m to about 13 m. As a result, the drawdown water volume can be reduced,
The amount of pool water in the suppression pool 7 can be reduced, and the entire volume of the containment vessel 8 can be reduced. The inner diameter of the containment vessel 8 is 3
It is possible to reduce the size from 8 m to about 37 m. Therefore, if this embodiment is adopted, the establishment of the raised suppression pool, which has conventionally tended to be considered to be economically poor, will be significantly higher. Fourth Embodiment FIG. 4 shows a main configuration of a reactor pressure vessel of a nuclear power plant according to a fourth embodiment of the present invention. In the present embodiment, a raised suppression pool type reactor containment that guides the main steam pipe 20 to the outside of the reactor containment vessel 8 through the suppression pool 7 located above the lowermost part of the reactor containment vessel 8. In a nuclear power plant having a vessel, a plurality of main steam isolation valves 21 and 22 provided in the main steam pipe 20 are all installed outside the reactor containment vessel 8, and a drawdown space below the reactor containment vessel 8 is provided. Is minimized. That is, in the present embodiment, the internal mechanism of the control rod drive mechanism 26 connected to the lower part of the reactor pressure vessel 1 is used for checking the soundness of the control rod drive mechanism (FMCRD) 26 at the time of periodic inspection. A structure in which the lower part of the control rod drive mechanism 26 is removed by pulling the reactor pressure vessel 1 upward in the reactor pressure vessel 1, and the reactor pressure vessel 1 is brought close to the lowermost part of the containment vessel 8. Has become. Conventionally, at the time of regular inspection, after the control rod (CR) and the control rod drive mechanism are separated from each other, the control rod drive mechanism is opened at its lower bottom and pulled out to the lower part of the reactor pressure vessel 1. I have. In that case, it is necessary to secure a space as a drawing margin below the reactor pressure vessel 1 further below the lower bottom of the control rod drive mechanism. Therefore, the height of the lower drywell was required to be about 11 m, despite the fact that the length of the control rod was 4 m. By employing a configuration in which the control rod drive mechanism 26 is pulled up at the time of regular inspection as in the present embodiment, it is possible to eliminate the need for a space below the reactor pressure vessel 1 as a withdrawal margin. As a result, the installation positions of the reactor pressure vessel 1 and the reactor core are lowered, and the effect of greatly improving the seismic resistance of the plant is obtained. Further, since the height of the reactor building can be reduced, the construction process can be significantly reduced. In the case of this embodiment, for example, the height of the containment vessel 8 can be reduced from the conventional 29.5 m to about 23 m. In the present embodiment, the wet well is not separately illustrated because it is considered separately as described in an embodiment described later. If the wet well is separately provided, the inner diameter of the lower dry well 4 is set to about 15 in this embodiment.
m. As a result, there is obtained an effect that a sufficient debris diffusion area can be secured at the time of a severe accident. However, the present invention is not limited to this, and it is of course possible to apply a combination with the first to third embodiments. When the suppression pool is of a raised type, an increase in the inner diameter of the lower dry well 4 causes an increase in drawdown water, which makes implementation difficult. In the present embodiment, the suppression pool is not a raised type. Further, in the present embodiment, a core catcher 27 using a heat-resistant member is provided on substantially the entire surface of the lower dry well 4. This makes it possible to significantly improve the soundness of the containment vessel at the time of a severe accident. Fifth Embodiment FIG. 5 shows a configuration of a nuclear power plant according to a fifth embodiment of the present invention. From the fifth embodiment to the eighth embodiment described in detail below, a safety system including two or more reactors, one wet well and a suppression pool in the same boiling water nuclear power plant, It is configured to be shared by the two or more nuclear reactors. In this embodiment, for example, a pair of reactors 1a and 1b having the configuration shown in the fourth embodiment is provided, and the two reactors 1a and 1b share one wet well 6 and suppression pool 7. Has become. That is, the dry well 2 is provided for each of the nuclear reactors 1a and 1b, but only one wet well 6 and the suppression pool 7 are provided. And
The dry wells 2 of both reactors 1a and 1b are connected to one suppression pool 7 by connecting vent pipes 28, respectively. Each connecting vent pipe 28 has a check valve or a rupture disk 29 described in a seventh embodiment to be described later. According to the present embodiment having the above configuration, one wet well 6 and suppression pool 7 are shared by two or more reactors 1a and 1ba, so that the total volume of the reactor containment vessel 8 is reduced. It is possible to greatly reduce. Sixth Embodiment FIG. 6 shows a configuration of a nuclear power plant according to a sixth embodiment of the present invention. In this embodiment, as in the fifth embodiment, a safety system including one wet well 6 and a suppression pool 7 is connected to two or more reactors 1a,
1b, a mid-well (MW) 30 is additionally provided above the wet well 6 for suppressing pressure. The dry wells 2 of the two reactors 1a and 1b are connected to the midwell 30 by connecting vent pipes 28, respectively, and the midwell 30 and the suppression pool 7 in the wet well 6 are connected to a large number of vertical vents. They are connected by a tube 31. That is, a large number of vertical vent pipes 31 are dropped from the midwell 30 into the suppression pool 7, and this
It is possible to share it in the plant. According to such a configuration, the number of vent pipes in the wet well 6 is one.
The plant is reduced to the amount corresponding to the wet well 6
Can function as a wet well for two plants with a design similar to that of a conventional containment vessel for almost one plant of the Mark II type reactor. Seventh Embodiment FIG. 7 shows a rupture disk 29 applied to the connecting vent pipe 28 of the containment vessel shown in FIGS. 5 and 6 as a seventh embodiment of the present invention. As shown in FIG. 7, the rupture disk 29 of the present embodiment is provided on the rupture disk main body 32 disposed inside the connecting vent pipe 28 and on the surface (back side) on the side of the fluid working direction a with respect to the rupture disk main body 32. And a pressure support mechanism 33. The pressure support mechanism 33 is configured by a support member having a rigid structure such as a mesh, for example.
The fluid pressure along the operating direction a is the rupture disc body 3
Notify 2 immediately. As a result, when pressure is applied in the operating direction a, only the pressure acts on the rupture disc main body 32, and the rupture disc main body 32 is immediately opened. On the other hand, when a back pressure is received from the reverse direction b, the rupture disc body 32
Is supported by the pressure support mechanism 33, so that the erroneous opening operation can be reliably prevented. Therefore, the rupture disk 29 of the healthy plant is
A malfunction occurs at the time of the coolant loss accident, and it is possible to prevent the influence of the coolant loss accident from affecting the healthy plant. Actually, about eight connecting vent pipes 28 are connected to the wet wells for each dry well, but in FIG. 5 and FIG. 6, only one connecting vent pipe 28 is shown for simplicity of explanation. That is, a total of about 16 vent pipes are connected from two plants, and these are further led into the suppression pool 7. Eighth Embodiment FIG. 8 shows a configuration of a nuclear power plant according to an eighth embodiment of the present invention. In the present embodiment, a plurality of, for example, two reactors 1a and 1b and one static safety system disposed therebetween, that is, the static containment cooling system 14,
The configuration includes an isolation condenser 15 and a gravity-fall type emergency core cooling system 13. The gravity drop type emergency core cooling system 13 is provided in a midwell 30 substantially similar to that shown in FIG.
The static containment cooling system 14 and the isolation condenser 15 are arranged above the midwell 30.
Then, the static containment vessel cooling system 14, the isolation condenser 15, and the gravity fall type emergency core cooling system 13
Consists of two reactors 1a, 1b
That is, the configuration is shared between the two plants. Other configurations are almost the same as those in FIG. According to such a configuration, the wet well 6 and the mid well 30 are shared between the two plants, and the static containment vessel cooling system 14, the isolation condenser 15, and the like are installed above the wet well 6, and the sixth embodiment. In addition to the functions and effects of the above, the static containment cooling system 1 has a large impact on the arrangement.
4. The pool of the isolation capacitors 15 can be efficiently arranged. In addition, since the gravity drop type emergency core cooling system 13 is arranged in the midwell 30, it is possible to completely eliminate the influence of the arrangement on the dry well 2 on each of the reactors 1a and 1b. Ninth Embodiment FIG. 9 is a plan view showing a configuration of a nuclear power plant according to a ninth embodiment of the present invention. In the present embodiment, not only the wet well 6 but also a plurality of, for example, two reactors 1a and 1b, that is, a dryer separator pit 41 which is conventionally installed for each plant, is shared between the two plants. ing. That is, as shown in FIG. 9, one dryer separator pit 41
It is arranged between the two reactors 1a and 1b. The dryer / separator pit 41 is arranged close to one wet well 6. The fuel pool 42
Are arranged outside each of the reactors 1a and 1b.
Below these fuel pools 42, a water pressure control unit (HCU), a reactor coolant purifier (CUW) and the like are provided. With such a configuration, it is possible to avoid unnecessary lengthening of the dimensions of the reactor building 40. The operating floor 43 of the reactor building is shared by two plants. Thereby, operability at the time of regular inspection can be greatly improved. In the case of ABWR, since the containment vessel of the reactor is extremely miniaturized, the operating floor of the reactor building also has only a minimum area required for operation. However, in this embodiment, the operating floor 4
3, it is possible to secure a wider operating floor area in practice, so that operability is greatly improved. Tenth Embodiment FIG. 10 shows a configuration of a nuclear power plant according to a tenth embodiment of the present invention. FIG. 10A shows a high-pressure core water injection system (HPCF) 51 and a low-pressure core water injection system (LPFL).
52, a line including a residual heat removal system (RHR) 53 (referred to as a front line for convenience). FIG. 10 (b) shows a reactor auxiliary equipment cooling system pump (RSW pump).
55 shows a line including a reactor water cooling pump (RCW pump) 56 and a reactor auxiliary equipment seawater cooling pump (RCW pump) 56 (referred to as a support line for convenience). In the present embodiment, FIG.
As shown in (a) and (b), as an emergency core cooling system, a high pressure core injection system (HPCF) 51, a low pressure core injection system (LPFL) 52, a residual heat removal system (RHR) 53, and a reactor accessory In a nuclear power plant equipped with a cooling system (RCW) 54, a high-pressure core water injection system 51, a low-pressure core water injection system 5
2. The front line including the residual heat removal system 53 and the emergency diesel power generation equipment (DG) 57 has a four-section configuration, while the support line having the piping configuration of the reactor auxiliary cooling system 54 has a two-loop configuration. And FIG.
As shown in FIG. 0 (b), two or more reactor accessory cooling system pumps (RS
W pump 55 and reactor auxiliary seawater cooling system pump (R
(CW pump) 56, and the emergency power supply for supplying power to them is divided into four sections. That is,
In this embodiment, the front line is completely
The system is divided, the support line is divided into two loops, all dynamic devices are installed at four or more, and an emergency core cooling system that enables online maintenance of the system is installed. When the emergency core cooling system configured as described above is shared between two plants, maintenance of the emergency core cooling system can be performed at an arbitrary time. Need to be implemented. In the conventional emergency core cooling system configuration, there are only two high-pressure core water injection systems (HPCF), and there is only one system, the reactor isolation cooling system (RCIC), so online maintenance of these systems is not possible. It was possible. In contrast, in the present embodiment, RCIC is deleted,
Instead, the high-pressure core water injection system (HPCF) has four systems, so that the effect of enabling online maintenance of all systems is obtained.

As described above, according to the present invention, by installing both main steam isolation valves outside the reactor containment vessel, the volume of the dry well can be reduced and the reactor containment vessel can be rationalized. At the same time, the peak pressure of the containment vessel at the time of the design standard coolant loss accident is reduced and safety is improved. In the raised suppression pool,
The volume of the lower part of the dry well, the volume of the suppression pool, and the like are reduced, and the entire reactor containment vessel can be made smaller. Also, the height of the lower drywell is reduced,
Since the height of the containment vessel and the reactor building is also reduced, the construction period of the plant can be significantly reduced. Further, since the wet well and the suppression pool are shared by two or more nuclear reactors, the reactor containment vessel can be made smaller than that provided for each plant. As a result, the size of the reactor building can be reduced, and the economics of the nuclear plant can be greatly improved. Nevertheless, since the operating floor of the reactor building is shared between plants, it is wider than when it is installed for each plant, and the operability during regular inspection is greatly improved. Furthermore, since the static safety system is shared between plants,
The placement effect can be greatly reduced. The emergency core cooling system composed of dynamic equipment can also be divided into four sections, including the high-pressure core water injection system, and online maintenance can be performed. Eliminates the need. All of the design standard accidents can be dealt with only by the safety system of the dynamic equipment, so there is no need to use the lower emergency core cooling system, which is a static safety system, at the time of the design standard accident. There is no need to install a pressure reducing valve. Since the lower type emergency core cooling system is installed as a pure backup of the dynamic safety system, it is sufficient if it can be injected into the core with sufficient time margin, so that extremely high reliability as the original static equipment is maintained It becomes possible. With the combined effects described above, the present invention improves the economics and safety of the nuclear power plant at the same time, and significantly improves the competitiveness of the nuclear power plant with other alternative power plants.

[Brief description of the drawings]

FIG. 1 is an explanatory view showing a first embodiment of a nuclear power plant according to the present invention.

FIG. 2 is an explanatory view showing a second embodiment of the nuclear power plant according to the present invention.

FIG. 3 is an explanatory view showing a third embodiment of the nuclear power plant according to the present invention.

FIG. 4 is an explanatory view showing a fourth embodiment of the nuclear power plant according to the present invention.

FIG. 5 is an explanatory view showing a fifth embodiment of the nuclear power plant according to the present invention.

FIG. 6 is an explanatory view showing a sixth embodiment of the nuclear power plant according to the present invention.

FIG. 7 is an explanatory view showing a rupture disk in a seventh embodiment of the nuclear power plant according to the present invention.

FIG. 8 is an explanatory view showing an eighth embodiment of the nuclear power plant according to the present invention.

FIG. 9 is an explanatory view showing a ninth embodiment of the nuclear power plant according to the present invention.

FIG. 10 (a) shows a first example of the nuclear power plant according to the present invention.
Explanatory drawing which shows the front line of the emergency core cooling system in 0th Embodiment, (b) is explanatory drawing which shows the support line of this emergency core cooling system.

FIG. 11 is an explanatory view showing an outline of a conventional containment vessel and the like of an ABWR.

FIG. 12 is an explanatory view showing an outline of a conventional SBWR reactor containment vessel and the like.

FIG. 13 is an explanatory diagram showing an outline of a raised S / P studied for a conventional next-generation ABWR.

[Explanation of symbols]

1. Reactor pressure vessel (RPV) 2. Dry well (D
W), 3 ... Upper dry well (UD), 4 ... Lower dry well (LD), 5 ... Vent tube, 6 ... Wet well (WW), 7 ... Suppression pool (S / P), 8 ... Reactor containment Vessel (PCV), 8a Guard pipe, 9 Primary piping, 10 Emergency core cooling system (ECCS), 11 Residual heat removal system (RHR), 12 RHR heat exchanger (RHR)
... Hx), 13 ... gravity fall type ECCS (GDCS), 1
4: Static containment cooling system (PCCS), 15: Isolation condenser (IC), 16: GDCS pool water, 17: PCCS heat exchange (PCCS: Hx), 18: P
CCS pool, 19 ... IC heat exchange (IC ... Hx), 20 ...
Main steam pipe (MS pipe), 21 ... 1st MSIV, 22 ... 2nd
MSIV, 23: Tunnel room, 24: ECCS room, 25
... PCV boundary, 26 ... Control rod drive mechanism (FMCRD),
27: core catcher, 28: connecting vent pipe, 29: rupture disk, 30: midwell (MW), 32: rupture disk body, 33: pressure support mechanism, 40: reactor building, 41: DS pit, 42
... fuel pool, 43 ... operating floor.

────────────────────────────────────────────────── ───

[Procedure amendment]

[Submission date] July 13, 1999 (1999.7.1)
3)

[Procedure amendment 1]

[Document name to be amended] Statement

[Correction target item name] Full text

[Correction method] Change

[Correction contents]

[Document Name] Statement

[Title of the Invention] Nuclear power plant

[Claims]

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a light water nuclear power plant (LWR), and more particularly to a nuclear power plant with a great improvement in economy and safety.

[0002]

2. Description of the Related Art A typical example of a conventional LWR is a boiling water reactor (BWR). Below,
The prior art will be described by taking the case of BWR as an example.

[0003] Each of the conventional BWRs incorporates one reactor pressure vessel (RPV) in a reactor containment vessel (PCV). As the latest BWR, an improved BWR (ABWR) is known. This ABWR
Is made of reinforced concrete called RCCV, and is the most sophisticated one currently in practical use.

FIG. 11 schematically shows a state of installation of a reactor containment vessel and the like of the ABWR. As shown in FIG. 11, the reactor pressure vessel 1 has a dry well (DW) 2
The dry well 2 is further divided into an upper dry well (UD) 3 and a lower dry well (LD) 4. The dry well 2 is connected to a suppression pool (S / S) in a wet well (WW) 6 by a vent pipe 5.
P) 7. The containment vessel 8 is thus mainly constituted by the dry well 2 and the wet well 6.

[0005] As a design standard accident of the containment vessel 8,
It is assumed that a primary pipe (main steam pipe) 9 connected to the reactor pressure vessel 1 breaks in the containment vessel 8. This is commonly referred to as a loss of coolant accident (LOCA). Even if a coolant loss accident occurs, the high-temperature steam released into the dry well 2 immediately passes through the vent pipe 5 and is condensed in the suppression pool 7, so that the internal pressure of the containment vessel 8 is sufficiently low. It is designed to be suppressed. In the conventional configuration, the main steam pipe 9 is provided with a first main steam isolation valve (first MS) located inside the containment vessel 8.
IV) 21 is provided.

[0006] An emergency core cooling system (ECCS) 10 using a pump 10a, which is a dynamic device, is installed to replenish the coolant that is lost in the event of a coolant loss accident. It is designed so that water can be pumped up and injected into the reactor pressure vessel 1.
In order to ensure reliability, a plurality of emergency core cooling systems, including those with different operating principles, are installed, and are designed to be extremely safe with multiplexing and diversity secured ( In the figure, only one system is shown for simplicity).

Further, in order to remove decay heat continuously generated from the reactor core even after the accident, a residual heat removal system (RHR) 1
1 is installed. This residual heat removal system uses a pump (RHR pump) 11a, which is also a dynamic device, to suck the suppression pool water, and feeds it to a heat exchanger (RHR).
It is designed so that the reactor containment vessel 8 can be cooled by circulating spray inside the reactor containment vessel 8 while being cooled by the heat exchanger 11b.

The residual heat removal system 11 is a safety system having a safety function as described above, but is also used as a cooling system when the reactor is normally shut down. Emergency diesel generator (emergency D / G) 12 as an emergency power source to operate dynamic safety systems such as the emergency core cooling system and residual heat removal system
Are installed (only one is shown in the figure). The residual heat removal system 11 shares the pump with the emergency core cooling system and forms a part of the emergency core cooling system.

On the other hand, although safety systems such as an emergency core cooling system and a residual heat removal system have not been put into practical use in recent years, all safety systems such as an emergency core cooling system and a residual heat removal system are performed only by static means, and dynamic devices such as pumps are changed from the safety system. Static safety furnace (SB
WR) has been proposed. Hereinafter, according to FIG.
The outline of R will be described. In FIG. 12, FIG.
11 are given the same reference numerals as in FIG. 11 and the description of the overlapping parts is omitted, and only the main parts will be described.

Referring to FIG. 12, a gravity-fall type emergency core cooling system (G
DCS) 13 is provided. A static containment cooling system (PCCS) 14 and an isolation condenser (IC) 15 are provided above the reactor containment vessel 8. When a coolant loss accident, which is a design standard accident, is assumed to occur, the point in which steam condenses in the suppression pool 7 is the same as in the case of the ABWR, but the coolant is supplied into the reactor pressure vessel 1. What we do is by gravity
The GDCS pool water 16 is naturally dropped.

In cooling the reactor containment vessel 8, steam and the like in the reactor containment vessel 8 are guided into the PCCS heat exchanger (PCCSHx) 17 by a slight differential pressure, and the PCCS pool 1 is cooled.
8 by cooling. Similarly, when the main steam isolation valve 21 is closed and the reactor is in an isolated state, the reactor steam in the reactor pressure vessel 1 is also cooled by I by buoyancy.
This is carried out by introducing into a C heat exchanger (ICHx) 19 for condensation.

As described above, in the case of the SBWR, when a certain abnormal situation occurs in the nuclear reactor, all the devices are statically operated by utilizing natural force. Therefore, these safety systems are simplified, reliability is greatly improved, and there is great expectation that an emergency power supply for operating them will not be required. Also in this example, a first main steam isolation valve (first MSIV) 21 is installed in the main steam pipe 9 inside the reactor containment vessel 8.

Further, in recent years, as a type of a powerful reactor containment vessel for the next-generation ABWR, a type having a suppression pool called a raised suppression pool has been studied (exhibition: Energy Review 19).
January 1998, p. 19).

An outline of the raised suppression pool will be described with reference to FIG. Note that FIG.
11 and 12 are the same as those in FIGS.
The same reference numerals are used to denote the same parts, and a description of overlapping parts will be omitted, and only main parts will be described.

As shown in FIG. 13, the dry well 2 forms an integral space, and there is no distinction between an upper dry well and a lower dry well. The main steam pipe (MS pipe) 20 hangs down in the dry well 2, and from the lower part of the dry well 2 through a tunnel room (MS tunnel room) 23 through the reactor containment vessel 8.
Guided outside. First main steam isolation valve 2 of main steam pipe 20
1 is provided on the dry well 2 side, and a second main steam isolation valve 2 is provided.
2 is disposed in the tunnel room 23.

Further, the suppression pool 7 is located considerably above the floor of the containment vessel 8, and a room (ECCS room) 24 therebelow can accommodate safety equipment such as an emergency core cooling system. . The gas phase portion of the wet well 6 extends to the upper end of the containment vessel 8 and has the same height as the top slab (top plate) of the dry well 2. In the drawing, reference numeral 25 denotes a reactor containment vessel boundary (PCV boundary).

In the raised suppression pool configured as described above, the gas phase space of the wet well 6 is large, and it is expected that an excellent improvement in safety that a rise in pressure at the time of an accident can be suppressed lower. Further, since safety equipment such as an emergency core cooling system can be housed in the ECCS room 24 in the lower part of the containment vessel 8, the floor area of the reactor building can be reduced, and a great improvement in arrangement can be expected.

[0018]

However, the above-described technique has various problems to be solved.

That is, the static safety furnace (S) shown in FIG.
In the case of a safety system of (BWR), a large amount of pool water for cooling is required, which has a large influence on the arrangement. For example, the PCCS pool 18 is about 2500 m ³ . For this reason, the reactor building cannot be downsized, which is an economical disadvantage. In addition, a gravity drop type emergency core cooling system (GDC
In step S) 13, there is only a water injection capability with a head having a slight difference in elevation, and it is necessary to rapidly depressurize the reactor pressure vessel 1 in the event of a coolant loss accident. For this reason, it is necessary to install a large number of pressure reducing valves. Considering the failure of the operation of the pressure reducing valves, the reliability of the gravity drop type emergency core cooling system (GDCS) 13 itself as an emergency core cooling system is extremely increased. It is difficult.

Although all safety systems are static,
The residual heat removal system for normal shutdown needs to leave the conventional dynamic equipment. Residual heat removal systems, including reactor auxiliary cooling system (RCW) and reactor auxiliary equipment seawater cooling system (RSW) (hereinafter referred to as "support system"), are expensive in safety systems, and eventually However, it has not led to the rationalization of safety systems. Furthermore, although an emergency power supply would have been unnecessary, in practice, some safety equipment had to be configured with dynamic devices, so some emergency power supply had to be left.

As described above, the SBWR is subject to the influence of the arrangement by the static safety system and the further rationalization of the dynamic system. In order to mitigate the effect of location, miniaturization of the PCV is also an important issue. Note that the SBWR has not been actually operated at the present time.

On the other hand, ABWR has already been put into actual operation,
Although it is recognized for its excellent safety, due to discussions on responding to severe accidents in recent years, in the design of the next-generation reactor next to ABWR, some static safety such as static containment cooling system etc. Additional systems are being considered to further enhance their safety. In the design of such a next-generation reactor, the influence on the layout of the reactor building becomes significant. To alleviate this, it is important to further reduce the volume of the containment vessel. Conversely, as the size of the containment vessel and reactor building of the ABWR has been reduced, the minimum required area of the operating floor of the reactor building has been secured, resulting in restrictions on operation during regular inspections. ing. Accordingly, it is important to increase the area of the operating floor of the reactor building while alleviating the influence of the layout of the entire building.

Although not shown, the present applicant has proposed a lower drywell separation system as a modified reactor containment vessel of the ABWR (for example, Japanese Patent Application No. 10-87411). ). According to this lower drywell separation method, it has become possible to significantly reduce the wetwell gas phase volume.
The upper drywell could not be made too small to secure the installation space for the main steam isolation valve, and there was some limit to the volume reduction of the reactor containment vessel as a whole. for that reason,
It is currently under consideration to reduce the volume of the containment vessel while keeping the space for installing the main steam isolation valve eccentric of the containment vessel. That is, even in the case of the lower drywell separation method, it is important to reduce the inner diameter of the upper drywell.

Further, in the case of the raised suppression pool shown in FIG. 13 which has been studied as a possible reactor containment type for the next-generation ABWR, the main steam pipe 20 passes through the lower part of the dry well 2. Therefore, the space under the dry well 2 becomes considerably larger than that of the conventional containment vessel. Since the suppression pool 7 is located at the top, when the emergency core cooling system is activated during an accident,
The suppression pool water stays in the lower space of the vast dry well 2 and does not return to the suppression pool 7. This is referred to as drawdown water, but to ensure from the beginning of the loss amount of the water source, it is necessary to increase the suppression pool water to conventionally about 3500 m ³ approximately doubles approximately 7000 m ^3.

As a result of considering the result, the volume of the wet well is increased, and eventually the volume of the entire containment vessel is increased, which leads to the deterioration of economy. Therefore, in the case of a raised suppression pool, it is necessary to reduce the volume of the space under the dry well 2 as much as possible. Can not be secured. Originally, the concept of a reactor containment vessel that was originally developed with an emphasis on the improvement of the placeability could result in a loss of its feasibility due to serious problems in the place. Therefore, in the case of a raised suppression pool, it is very important to make the space under the drywell more compact without sacrificing the arrangement of such main devices.

The present invention has been made in view of such circumstances, and a compact nuclear power plant that can simultaneously improve economic efficiency and safety, and can further significantly improve the competitiveness of nuclear power generation with other alternative power generation facilities. The purpose is to provide power generation equipment.

[0027]

According to one aspect of the present invention, a reactor vessel is housed in a reactor containment vessel, and a main steam pipe connected to the reactor vessel is provided. In the nuclear power plant configured to penetrate through the reactor containment and lead out to the outside, a plurality of main steam isolation valves provided in the main steam pipe are all installed outside the reactor containment. And

That is, with respect to the conventional two main steam isolation valves, the installation locations are separated such that the first main steam isolation valve is inside the reactor containment vessel and the second main steam isolation valve is outside the reactor containment vessel. In the present invention, both of the two valves are installed outside the reactor containment vessel. Therefore, according to the first aspect of the present invention, since the main steam isolation valve is released from extremely severe environmental conditions such as temperature, pressure, and radioactivity that are exposed in a severe accident, for example, Can be increased.

Conventionally, the main steam isolation valve provided in the containment vessel is the largest factor determining the inner diameter of the dry well except for the reactor pressure vessel body.
By eliminating this, the inner diameter of the dry well can be greatly reduced. As the inner diameter of the drywell becomes smaller, the volume of the drywell becomes smaller.This not only greatly reduces the volume of the reactor but also reduces the pressure in the reactor containment vessel in the event of a loss of coolant. Has a positive effect. That is, when the dry well volume is reduced, the amount of non-condensable gas such as nitrogen existing before the accident can be reduced, and in the event of a coolant loss accident, nitrogen or the like moves to the wet well gas phase and is compressed. This is because the rise in the internal pressure of the containment vessel due to the above can be further reduced.

The invention according to claim 2 is a raised suppression pool type reactor in which the main steam pipe is guided to the outside of the reactor containment vessel through the suppression pool located above the lowermost part of the containment vessel. In a nuclear power plant having a containment vessel, a plurality of main steam isolation valves provided in the main steam pipe are all installed outside the reactor containment vessel, and a drawdown space below the reactor containment vessel is minimized. It is characterized by.

According to the present invention, two main steam isolation valves are installed in a tunnel room on the outer peripheral side of the dry well.
When applied to a raised suppression pool, the space under the drywell can be made sufficiently small without sacrificing the arrangement of the main steam isolation valve. As a result, the drawdown water amount is greatly reduced, and the suppression pool water amount is also reduced, so that the wet well volume is reduced, the reactor containment volume is reduced, and the cost of the reactor containment can be reduced.

The invention according to claim 3 is the invention according to claim 1 or 2
The nuclear power plant described above is characterized in that, of the main steam pipe, a pipe portion from a penetrating portion of the containment vessel to a main steam isolation valve disposed outside the main vessel has a double structure.

According to the present invention, when two main steam isolation valves are installed outside the reactor containment vessel or in the tunnel room,
Since the main steam pipe between the second main steam isolation valve and the reactor containment vessel has a double pipe structure, even if the main steam pipe up to the second main steam isolation valve is broken, it has a double structure. In addition, it is possible to eliminate the influence of the pipe break accident. This allows
Even when the two main steam isolation valves are installed outside the containment vessel, sufficient safety can be ensured.

According to a fourth aspect of the present invention, in the boiling water nuclear power plant, the control rod drive mechanism connected to the lower part of the reactor pressure vessel is drawn upward in the reactor pressure vessel. The lower space of the control rod drive mechanism is deleted, and the reactor pressure vessel is made to approach the lowermost part of the containment vessel.

That is, according to the present invention, the control rod drive mechanism (FMCRD) which is installed at the lower part of the reactor pressure vessel, which has been pulled down at the time of maintenance or the like in the past, is changed to the reactor pressure vessel. By pulling out the inside of the vessel, that is, the upper side, most of the lower dry well space, for example, about 11 m, which was conventionally installed at the lower part of the reactor pressure vessel, is deleted, and the reactor pressure vessel is reduced to the lower floor of the containment vessel as much as possible. Set up close to the surface.

Therefore, according to the present invention, since the position of the reactor pressure vessel, which is the largest heavy object, shifts significantly downward, the seismic conditions of the core fuel are greatly relaxed, and the seismic resistance of the reactor plant is greatly improved. Can be improved. Furthermore, while the height of the reactor containment vessel was conventionally determined as the sum of the reactor pressure vessel height and the lower drywell height, according to the present invention, the lower drywell height is reduced by half. Thereby, the height of the reactor containment vessel can be significantly reduced. As a result, the reactor containment vessel can be rationalized, and further, the reactor building itself can be rationalized. By reducing the height of the reactor building, for example, the reactor building can be reduced by one level, whereby the plant construction process can be greatly shortened. For example, it is possible to reduce the world's shortest 48-month process to the level of a 36-month process even today.

According to a fifth aspect of the present invention, in the boiling water nuclear power plant, two or more nuclear reactors are provided in the same facility, and a safety system comprising one wet well and a suppression pool is provided for the two or more nuclear reactors. It is characterized in that it is shared by furnaces.

According to the present invention, in a reactor containment vessel of a boiling water nuclear power plant, a wet well and a suppression pool are shared by two or more reactors, and therefore, conventionally, each reactor is provided for each reactor. Only one wet well and one suppression pool are required, so that the entire volume of the containment vessel can be significantly reduced and the arrangement inside the reactor building can be greatly eased.

According to a sixth aspect of the present invention, in the nuclear power plant according to the fifth aspect, a midwell for suppressing pressure is newly provided above the wet well, and the midwell and a suppression pool in the wet well are provided. Are connected by a vent pipe.

According to the present invention, a common space called a midwell is newly provided, and this space and two or more drywells are connected by a vent pipe. The connection to the pool eliminates the need to separately connect vent tubes from two or more dry wells to wet wells. Thereby, only the vent pipe for one plant needs to be connected from the midwell to the wet well. That is, if the vent pipes for the two plants are connected directly to the wet wells independently of each other, the number of vent pipes guided into the suppression pool becomes considerably large, and the wet wells need to be enlarged. In the present invention, this can be avoided.

The invention according to claim 7 is the invention according to claim 5 or 6.
In the nuclear power plant described above, a dry well in each nuclear reactor and one shared wet well or midwell are connected by a connecting vent pipe, and a check valve or a rear side is connected to the connecting vent pipe. Is provided with a rupture disk having a pressure support mechanism.

According to the present invention, the check valve or the rupture disk provided with the pressure support mechanism on the back surface is provided in the vent pipe so that the dry well and the wet well do not come into direct contact with each other at normal times. Even if the dry well is opened during the regular inspection of the reactor, it is possible to prevent any safety problems from occurring.

Since the rear side of the rupture disk is supported by the pressure indicating mechanism, there is no risk of malfunction even when a back pressure is applied.

When a loss of coolant accident occurs in one reactor, the check valve or the rupture disk of the connecting vent pipe installed between the reactor and the dry well of the healthy side may malfunction and cause an accident. It is possible to avoid the effect of the above on the reactor side on the healthy side.

The invention according to claim 8 is the light water nuclear power plant, wherein two or more reactors are provided in the same facility,
One static safety system is shared by the two or more nuclear reactors.

Here, the static safety system includes a static containment vessel cooling system, an isolation condenser, a gravity drop type emergency core cooling system, and the like.

According to the present invention, the influence of the arrangement of the static safety system, which has conventionally had a great influence, can be greatly reduced. The present invention can be applied to all light water nuclear power plants using a static safety system.

According to a ninth aspect of the present invention, in a light water nuclear power plant, two or more nuclear reactors are provided in the same facility,
It is characterized in that a dryer, a separator, and a pit are shared by the two or more nuclear reactors.

Here, the dryer separator pit (D
S pit) refers to storage pits such as a steam dryer and a steam separator.

According to the present invention, it is not necessary to install a dryer / separator pit for each plant, so that the operating floor can be used more widely. Further, it is not necessary to unnecessarily increase the floor area of the reactor building.
Conventionally, the fuel pool and the dryer / separator pit are arranged in a straight line, and if these are arranged in series for two plants, the size of the reactor building becomes very long. According to the present invention, this is avoided. It becomes possible. That is, since the dryer / separator pit is used at the time of refueling of the plant, it is not necessary to install the dryer / separator pit for each plant unless simultaneous inspection of the two plants is performed. In addition, in the case of a request for simultaneous inspection of two plants,
Only the steam-water separator that needs to be stored underwater can be handled by securing the capacity of the dryer / separator pit so that two plants can be stored.

According to a tenth aspect of the present invention, there is provided a nuclear power plant comprising a high-pressure core cooling system, a low-pressure core cooling system, a residual heat removal system, and a reactor auxiliary cooling system as an emergency core cooling system. The cooling system, the low-pressure core cooling system, and the residual heat removal system have a four-section configuration, while the piping of the reactor auxiliary cooling system has a two-loop configuration.

According to an eleventh aspect of the present invention, in the nuclear power plant according to the tenth aspect, two or more reactor accessory cooling system pumps and reactor accessories are provided in the piping of each loop of the reactor accessory cooling system. A seawater cooling system is provided, and an emergency power supply for supplying power to the seawater cooling system is divided into four sections.

According to the tenth and eleventh aspects of the present invention, all the front lines of the emergency core cooling system including the high pressure core cooling system are divided into four sections. All of the support lines can be maintained online.

A twelfth aspect of the present invention is characterized in that the emergency core cooling system according to the tenth or eleventh aspect is configured to be shared by two or more nuclear reactors.

According to the present invention, all parts of the emergency core cooling system can be maintained on-line, so that the maintenance of the emergency core cooling system can be performed at any time without simultaneous inspection of two plants. Becomes

The invention according to claim 13 is the invention according to claims 1 to 3
The configuration of the main steam pipe according to any one of the above, and the configuration of the reactor pressure vessel and the containment vessel according to claim 4,
It is characterized by having both.

According to a fourteenth aspect of the present invention, there are provided a plurality of nuclear reactors having the configuration according to the thirteenth aspect, and the safety system or the dryer separator according to any one of the fifth to ninth aspects is provided for these reactors. -It is characterized by the combination of pit configurations.

[0058]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of a nuclear power plant according to the present invention will be described below with reference to FIGS.

First Embodiment FIG. 1 shows a configuration of a reactor containment vessel of a nuclear power plant and its periphery according to a first embodiment of the present invention.

In this embodiment, as shown in FIG. 1, a reactor pressure vessel 1 is installed in a dry well 2, and the dry well 2 is further divided into an upper dry well 3 and a lower dry well 4. The dry well 2 is connected to a suppression pool 7 in the wet well 6 by a vent pipe 5. The reactor containment vessel 8 is thus of a lower drywell separation type.

In such a configuration, in the present embodiment, the main steam pipe 9 connected to the reactor pressure vessel 1 passes through the reactor containment vessel 8 and is led out to the outside. , A plurality of main steam isolation valves, for example, first and second two main steam isolation valves 21 and 22 are all disposed outside the reactor containment vessel 8.

According to such a configuration, the first main steam isolation valve 21 conventionally disposed in the upper dry well 3 is disposed outside the reactor containment vessel 8 so that the first main steam isolation valve 21 is disposed in the upper dry well 3. The space for the main steam isolation valve 21 is not required, so that the inner diameter of the upper dry well 8 can be greatly reduced. For example, the inner diameter of the dry well can be reduced from the current 29 m to about 20 m as shown in FIG.

Further, when the volume of the dry well is reduced by such a configuration, the amount of nitrogen which is a non-condensable gas existing in the dry well is reduced, and the nitrogen gas contained in the wet well side in the event of a loss of coolant. Will be greatly reduced. As a result, the effect of improving safety that the peak pressure of the containment vessel 8 at the time of the coolant loss accident can be reduced can be obtained.

In the prior art, the first main steam isolation valve inside the containment vessel was exposed to severe environmental conditions in the containment vessel at the time of a severe accident or the like. This problem can be fundamentally solved by installing the one-handed steam isolation valve 21 outside the containment vessel 8.

Furthermore, the accessibility to the main steam isolation valve 21 is improved, and the dose of the operator exposed in the containment vessel 8 during maintenance can be greatly reduced. As shown in FIG. 1, in the present embodiment, the inner diameter of the upper dry well 3 is about 20 m and the inner diameter of the wet well 6 is about 26 m. This is for securing the pool water volume of the pool 7.

As a modified example of the present embodiment, when the above-described main steam isolation valve arrangement is applied to a reactor containment vessel of a type in which a wet well is separately provided, the reactor containment vessel is not shown. Can be made unitary as a whole to match the upper drywell inner diameter. In that case, the shape of the containment vessel can be made a uniform cylindrical shape.

The numerical values shown in FIG. 1 are merely examples, and can be variously changed. The number of the main steam isolation valves 21 and 22 provided outside the reactor containment vessel 8 is also shown in FIG. It is possible to set it to an arbitrary number of 2 or more. As described above, the dimensions of each part of the reactor containment vessel 8 and the number of valves and other parts are not necessarily limited to those shown in the drawings and the like, and various changes can be made in the following embodiments. is there.

Second Embodiment FIG. 2 shows a reactor containment vessel of a nuclear power plant and its peripheral configuration according to a second embodiment of the present invention.

Also in this embodiment, as in the first embodiment, a plurality of, for example, first and second
The two main steam isolation valves 21 and 22 are arranged outside the reactor containment vessel 8.

Further, in this embodiment, a portion of the main steam pipe 9 from a portion penetrating the reactor containment vessel 8 to a first main steam isolation valve 21 installed outside the reactor containment vessel 8, A portion from the first main steam isolation valve 21 to the second main steam isolation valve 22 has a double pipe structure covered with a guard pipe 8a. Other configurations are the same as those of the first embodiment, and therefore, the same reference numerals as in FIG. 1 are assigned to FIG. 2 and the description is omitted.

According to the structure of the present embodiment, the following effects can be obtained in addition to the effects of the first embodiment.

That is, conventionally, since the main steam pipe between the main steam isolation valves is the boundary of the containment vessel, the assumption of the breakage of the pipe has not been required in the safety design. However, considering the piping breakage at this portion, even if the second main steam isolation valve outside the reactor containment vessel is closed as designed, its isolation function is lost. Therefore, even in this case, the first main steam isolation valve has been installed inside the reactor containment vessel so that the containment vessel can be isolated.

On the other hand, in the present embodiment, the portion of the main steam pipe 8 up to the second main steam isolation valve 22 is the guard pipe 8
a, the first main steam isolation valve 2 is protected even if the pipe is broken at this point.
It becomes possible to eliminate the necessity of installing the inside 1 of the containment vessel 8.

Third Embodiment FIG. 3 shows a configuration of a containment vessel of a nuclear power plant according to a third embodiment of the present invention.

This embodiment is applied to the above-mentioned raised suppression pool type reactor containment vessel, and the overall configuration is substantially the same as that shown in FIG. 13, so that FIG. The description is omitted, and only the main configuration will be described.

As shown in FIG. 3, in this embodiment, the main steam pipe 9 connected to the reactor pressure vessel 1 is led out of the containment vessel 8. In this embodiment, both the main steam isolation valves 21 and 22 are installed in a tunnel room 23 outside the reactor containment vessel 8, and the piping upstream of the main steam isolation valves 21 and 22 in the tunnel room 23 is provided. A portion is covered with a guard pipe 9b to form a double piping structure.

According to such a configuration, since both the first main steam isolation valve 2 are disposed outside the reactor containment vessel 8, the lower space of the dry well 2 is minimized by that much. The size has become. For example, under such a configuration of the present embodiment, the inner diameter of the lower portion of the dry well 2 can be reduced from the current 17 m to about 13 m.

As a result, since the drawdown water amount can be reduced, the pool water amount of the suppression pool 7 can be reduced, and the entire volume of the reactor containment vessel 8 can be reduced. The inner diameter of the containment vessel 8 can be reduced from the current 38 m to about 37 m. Therefore, if this embodiment is adopted, the establishment of the raised suppression pool, which has conventionally tended to be considered to be economically poor, will be significantly higher.

Fourth Embodiment FIG. 4 shows a configuration of a main part of a reactor pressure vessel of a nuclear power plant according to a fourth embodiment of the present invention.

In this embodiment, a raised suppression pool type is used, in which the main steam pipe 20 is guided to the outside of the reactor containment vessel 8 through the suppression pool 7 located above the lowermost part of the containment vessel 8. In a nuclear power plant having a containment vessel, a plurality of main steam isolation valves 21 and 22 provided in the main steam pipe 20 are all installed outside the containment vessel 8, and the
The lower draw-down space is minimized.

That is, in this embodiment, when the soundness of the control rod drive mechanism (FMCRD) 26 is checked at the time of the periodic inspection, the inside of the control rod drive mechanism 26 connected to the lower part of the reactor pressure vessel 1 is checked. The mechanism is configured to be pulled out upward in the reactor pressure vessel 1, whereby the space under the control rod drive mechanism 26 is deleted, and the reactor pressure vessel 1 is brought close to the bottom of the reactor containment vessel 8. Configuration.

Conventionally, at the time of regular inspection, after the control rod (CR) and the control rod drive mechanism are separated from each other, the control rod drive mechanism is opened at its lower bottom and pulled out to the lower part of the reactor pressure vessel 1. Has taken. In that case, it is necessary to secure a space as a drawing margin below the reactor pressure vessel 1 further below the lower bottom of the control rod drive mechanism. Therefore, the height of the lower drywell was required to be about 11 m, despite the fact that the length of the control rod was 4 m.

As in the present embodiment, when the control rod drive mechanism 26 is pulled up at the time of regular inspection, a space as a draw-out margin below the reactor pressure vessel 1 can be eliminated. As a result, the installation positions of the reactor pressure vessel 1 and the reactor core are lowered, and the effect of greatly improving the seismic resistance of the plant is obtained.

Further, since the height of the reactor building can be reduced, the construction process can be greatly reduced. In the case of this embodiment, for example, the height of the containment vessel 8 can be reduced from the conventional 29.5 m to about 23 m. In the present embodiment, the wet well is not separately illustrated because it is considered separately as described in an embodiment described later. If the wet well is separately provided, the inner diameter of the lower dry well 4 is set to about 15 in this embodiment.
m. As a result, there is obtained an effect that a sufficient debris diffusion area can be secured at the time of a severe accident. However, the present invention is not limited to this, and it is of course possible to apply a combination with the first to third embodiments.

When the suppression pool is of a raised type, an increase in the inner diameter of the lower dry well 4 causes an increase in drawdown water, which makes implementation difficult. In the present embodiment, the suppression pool is not a raised type.

Further, in this embodiment, the core catcher 2 using a heat-resistant member is provided on almost the entire surface of the lower dry well 4.
7 are installed. This makes it possible to significantly improve the soundness of the containment vessel at the time of a severe accident.

Fifth Embodiment FIG. 5 shows the configuration of a nuclear power plant according to a fifth embodiment of the present invention.

From the fifth embodiment to the eighth embodiment, which will be described in detail below, the same boiling water nuclear power plant
A safety system comprising one or more nuclear reactors and one wet well and a suppression pool is shared by the two or more nuclear reactors.

In the present embodiment, for example, a pair of reactors 1a and 1b having the configuration shown in the fourth embodiment is provided, and one wet well 6 and suppression pool 7 are shared by the two reactors 1a and 1b. Configuration. That is, the dry well 2 is provided for each of the nuclear reactors 1a and 1b, but only one wet well 6 and the suppression pool 7 are provided.

The dry wells 2 of the two reactors 1a and 1b are connected to one suppression pool 7 by connecting vent pipes 28, respectively. Each connecting vent pipe 28 has a check valve or a rupture disk 29 described in a seventh embodiment to be described later.

According to the present embodiment having the above configuration,
With the configuration in which one wet well 6 and the suppression pool 7 are shared by two or more reactors 1a and 1ba, the total volume of the reactor containment vessel 8 can be significantly reduced.

Sixth Embodiment FIG. 6 shows the configuration of a nuclear power plant according to a sixth embodiment of the present invention.

In the present embodiment, as in the fifth embodiment, 1
The safety system including one wet well 6 and the suppression pool 7 is configured to be shared by two or more reactors 1a and 1b.
On the upper side, a mid-well (M
W) 30 is provided. The dry wells 2 of the two reactors 1a and 1b are connected to the midwell 30 by connecting vent pipes 28, respectively, and the midwell 30 and the suppression pool 7 in the wet well 6 are connected to a large number of vertical vents. They are connected by a tube 31.

That is, a large number of vertical vent pipes 31 are dropped from the midwell 30 into the suppression pool 7, and this group of vent pipes 31 can be shared by two plants.

According to such a configuration, the number of vent pipes in the wet well 6 is reduced to one plant, whereby the wet well 6 is connected to the reactor containment vessel of almost one plant of the conventional Mark II reactor. With a similar design,
It is possible to function as a wet well for two plants.

Seventh Embodiment FIG. 7 shows the structure of a rupture disk 29 applied to the connecting vent pipe 28 of the reactor containment vessel shown in FIGS. 5 and 6 as a seventh embodiment of the present invention. .

The rupture disk 29 of this embodiment is
As shown in FIG. 7, the rupture disk main body 32 is disposed inside the connecting vent pipe 28, and the pressure support mechanism 33 is provided on the surface (rear side) on the side of the fluid working direction a with respect to the rupture disk main body 32. I have.

The pressure support mechanism 33 is constituted by a support member having a rigid structure such as a mesh, for example. Fluid pressure along the operating direction a is immediately transmitted to the rupture disk main body 32 by the mesh of the pressure support mechanism 33. . As a result, when pressure is applied in the operating direction a, only the pressure acts on the rupture disc main body 32, and the rupture disc main body 32 is immediately opened.

On the other hand, when a back pressure is applied in the reverse direction b, the rupture disk main body 32 is supported by the pressure support mechanism 33, so that the erroneous opening operation can be reliably prevented. Therefore, it is possible to prevent the rupture disk 29 of the healthy plant from malfunctioning at the time of the coolant loss accident, and to prevent the influence of the coolant loss accident from affecting the healthy plant.

Note that actually the connecting vent pipe 2
8 is connected to the wet well for each dry well, and about 8 wires are connected to the wet well.
Only one is displayed. That is, a total of about 16 vent pipes are connected from two plants, and these are further led into the suppression pool 7.

Eighth Embodiment FIG. 8 shows the configuration of a nuclear power plant according to an eighth embodiment of the present invention.

In this embodiment, a plurality of, for example, two reactors 1a and 1b, and one static safety system disposed therebetween, that is, a static containment cooling system 14, an isolation condenser 15, and a gravity It is configured to include a drop-type emergency core cooling system 13. The gravity-fall type emergency core cooling system 13 is provided in a midwell 30 substantially similar to that shown in FIG. 6, and the static containment vessel cooling system 14 and the isolation condenser 15 are provided in the midwell 30. It is arranged above 30.

One static safety system consisting of the static containment vessel cooling system 14, the isolation condenser 15, and the gravity drop type emergency core cooling system 13 is provided between the two reactors 1a and 1b, that is, between the two plants. It is configured to be shared. Other configurations are almost the same as those in FIG.

According to such a configuration, the wet well 6 and the midwell 30 are shared between the two plants, and the static containment vessel cooling system 14, the isolation condenser 15, and the like are provided above the wet well 6 and the mid well 30, so that the second plant can be used. In addition to the functions and effects of the sixth embodiment, it is possible to efficiently arrange the pool of the static containment vessel cooling system 14 and the isolation condenser 15 having a large impact on the arrangement. In addition, since the gravity drop type emergency core cooling system 13 is arranged in the midwell 30, it is possible to completely eliminate the influence of the arrangement on the dry well 2 on each of the reactors 1a and 1b.

Ninth Embodiment FIG. 9 is a plan view showing the configuration of a nuclear power plant according to a ninth embodiment of the present invention.

In the present embodiment, not only the wet well 6 but also a plurality of, for example, two reactors 1a and 1b, that is, the dryer / separator pits 41 which are conventionally installed for each plant are shared by the two plants. It has a configuration.

That is, as shown in FIG. 9, one dryer separator pit 41 includes two reactors 1a,
1b. The dryer / separator pit 41 is arranged close to one wet well 6. The fuel pool 42 is provided for each of the reactors 1a, 1
One pair is arranged outside b. Below these fuel pools 42, a water pressure control unit (HCU), a reactor coolant purifier (CUW) and the like are provided.

With such a configuration, it is possible to avoid an unnecessary increase in the size of the reactor building 40. The operating floor 43 of the reactor building is shared by two plants. Thereby, operability at the time of regular inspection can be greatly improved. In the case of ABWR, since the containment vessel of the reactor is extremely miniaturized, the operating floor of the reactor building also has only a minimum area required for operation. However, in this embodiment, the operating floor 4
3, it is possible to secure a wider operating floor area in practice, so that operability is greatly improved.

Tenth Embodiment FIG. 10 shows the configuration of a nuclear power plant according to a tenth embodiment of the present invention. FIG. 10A shows a high-pressure core water injection system (HPCF) 51 and a low-pressure core water injection system (LPFL).
52, a line including a residual heat removal system (RHR) 53 (referred to as a front line for convenience). FIG. 10 (b) shows a reactor auxiliary equipment cooling system pump (RSW pump).
55 shows a line including a reactor water cooling pump (RCW pump) 56 and a reactor auxiliary equipment seawater cooling pump (RCW pump) 56 (referred to as a support line for convenience).

In the present embodiment, as shown in FIGS. 10A and 10B, as an emergency core cooling system, a high pressure core water injection system (HPCF) 51, a low pressure core water injection system (LPFL) 52,
In a nuclear power plant equipped with a residual heat removal system (RHR) 53 and a reactor auxiliary cooling system (RCW) 54, a high-pressure core water injection system 51, a low-pressure core water injection system 52, a residual heat removal system 5
3, and the front line including the emergency diesel power generator (DG) 57 is divided into four sections, while the support line having the piping configuration of the reactor auxiliary cooling system 54 is changed to two sections.
It has a loop configuration.

As shown in FIG. 10B, two or more reactor accessory cooling system pumps (RSW pumps) 55 and reactor accessory equipment are provided in the piping of each loop of the reactor accessory cooling system 54. A seawater cooling system pump (RCW pump) 56 is provided, and an emergency power supply for supplying power to these pumps is provided in four sections.

That is, in the present embodiment,
Completely divided the front line into 4 sections and 2 support lines
The system is looped, all the dynamic devices are installed at least 4 units, and an emergency core cooling system that enables online maintenance of the system is installed.

When the emergency core cooling system configured as described above is shared between two plants, maintenance of the emergency core cooling system can be performed at an arbitrary time. Eliminates the need to perform system maintenance. In the conventional emergency core cooling system configuration, there are only two systems of high pressure core injection system (HPCF), and there is only one system for reactor isolation cooling system (RCIC)
The on-line maintenance of these lines was not possible due to the presence of

On the other hand, in this embodiment, since the RCIC is deleted and the high pressure core water injection system (HPCF) is replaced with four systems, the effect that all systems can be maintained online can be obtained.

[0115]

As described above, according to the present invention, by installing both main steam isolation valves outside the reactor containment vessel, the volume of the dry well can be reduced and the reactor containment vessel can be rationalized. At the same time, the peak pressure of the containment vessel at the time of the design standard coolant loss accident is reduced and safety is improved. In the raised suppression pool,
The volume of the lower part of the dry well, the volume of the suppression pool, and the like are reduced, and the entire reactor containment vessel can be made smaller. Also, the height of the lower drywell is reduced,
Since the height of the containment vessel and the reactor building is also reduced, the construction period of the plant can be significantly reduced. Further, since the wet well and the suppression pool are shared by two or more nuclear reactors, the reactor containment vessel can be made smaller than that provided for each plant. As a result, the size of the reactor building can be reduced, and the economics of the nuclear plant can be greatly improved. Nevertheless, since the operating floor of the reactor building is shared between plants, it is wider than when it is installed for each plant, and the operability during regular inspection is greatly improved. Furthermore, since the static safety system is shared between plants,
The placement effect can be greatly reduced. The emergency core cooling system composed of dynamic equipment can also be divided into four sections, including the high-pressure core water injection system, and online maintenance can be performed. Eliminates the need. All of the design standard accidents can be dealt with only by the safety system of the dynamic equipment, so there is no need to use the lower emergency core cooling system, which is a static safety system, at the time of the design standard accident. There is no need to install a pressure reducing valve. Since the lower type emergency core cooling system is installed as a pure backup of the dynamic safety system, it is sufficient if it can be injected into the core with sufficient time margin, so that extremely high reliability as the original static equipment is maintained It becomes possible.

With the above combined effects, the present invention simultaneously improves the economics and safety of a nuclear power plant,
This has the effect of significantly improving the competitiveness of nuclear power with other alternative power generation facilities.

[Brief description of the drawings]

FIG. 1 is an explanatory view showing a first embodiment of a nuclear power plant according to the present invention.

FIG. 2 is an explanatory view showing a second embodiment of the nuclear power plant according to the present invention.

FIG. 3 is an explanatory view showing a third embodiment of the nuclear power plant according to the present invention.

FIG. 4 is an explanatory view showing a fourth embodiment of the nuclear power plant according to the present invention.

FIG. 5 is an explanatory view showing a fifth embodiment of the nuclear power plant according to the present invention.

FIG. 6 is an explanatory view showing a sixth embodiment of the nuclear power plant according to the present invention.

FIG. 7 is an explanatory view showing a rupture disk in a seventh embodiment of the nuclear power plant according to the present invention.

FIG. 8 is an explanatory view showing an eighth embodiment of the nuclear power plant according to the present invention.

FIG. 9 is an explanatory view showing a ninth embodiment of the nuclear power plant according to the present invention.

FIG. 10 (a) shows a first example of the nuclear power plant according to the present invention.
Explanatory drawing which shows the front line of the emergency core cooling system in 0th Embodiment, (b) is explanatory drawing which shows the support line of this emergency core cooling system.

FIG. 11 is an explanatory view showing an outline of a conventional containment vessel and the like of an ABWR.

FIG. 12 is an explanatory view showing an outline of a conventional SBWR reactor containment vessel and the like.

FIG. 13 is an explanatory diagram showing an outline of a raised S / P studied for a conventional next-generation ABWR.

[Description of Signs] 1 ... Reactor Pressure Vessel (RPV) 2 ... Dry Well (D
W), 3 ... Upper dry well (UD), 4 ... Lower dry well (LD), 5 ... Vent tube, 6 ... Wet well (WW), 7 ... Suppression pool (S / P), 8 ... Reactor containment Vessel (PCV), 8a Guard pipe, 9 Primary piping, 10 Emergency core cooling system (ECCS), 11 Residual heat removal system (RHR), 12 RHR heat exchanger (RHR)
... Hx), 13 ... gravity fall type ECCS (GDCS), 1
4: Static containment cooling system (PCCS), 15: Isolation condenser (IC), 16: GDCS pool water, 17: PCCS heat exchange (PCCS: Hx), 18: P
CCS pool, 19 ... IC heat exchange (IC ... Hx), 20 ...
Main steam pipe (MS pipe), 21 ... 1st MSIV, 22 ... 2nd
MSIV, 23: Tunnel room, 24: ECCS room, 25
... PCV boundary, 26 ... Control rod drive mechanism (FMCRD),
27: core catcher, 28: connecting vent pipe, 29: rupture disk, 30: midwell (MW), 32: rupture disk body, 33: pressure support mechanism, 40: reactor building, 41: DS pit, 42
... fuel pool, 43 ... operating floor.

Claims (14)

[Claims] 1. A nuclear power plant having a structure in which a reactor vessel is housed in a reactor vessel and a main steam pipe connected to the reactor vessel is led out through the reactor vessel. A nuclear power plant, wherein a plurality of main steam isolation valves provided in the main steam pipe are all installed outside the reactor containment vessel. 2. A nuclear power plant having a raised suppression pool type reactor containment vessel for guiding a main steam pipe to the outside of the reactor containment vessel through a suppression pool located above a lowermost part of the reactor containment vessel. In the facility, the plurality of main steam isolation valves provided in the main steam pipe are all installed outside the reactor containment vessel, and the drawdown space at the lower part of the reactor containment vessel is minimized. Power generation equipment. 3. The nuclear power plant according to claim 1, wherein a portion of the main steam pipe from a penetration portion of the containment vessel to a main steam isolation valve disposed outside the main steam pipe has a double structure. A nuclear power plant characterized by the following. 4. In the boiling water nuclear power plant, a control rod drive mechanism connected to a lower part of the reactor pressure vessel is drawn upward in the reactor pressure vessel, so that the control rod drive mechanism is A nuclear power plant, wherein a lower space is removed, and the reactor pressure vessel is made to approach the lowermost part of the containment vessel. 5. A boiling water nuclear power plant, comprising two or more reactors in the same facility, and a safety system comprising one wet well and a suppression pool.
A nuclear power plant characterized in that it is configured to be shared by two or more nuclear reactors. 6. The nuclear power plant according to claim 5, further comprising a new mid-well for suppressing pressure above the wet well, wherein the mid-well and the suppression pool in the wet well are connected by a vent pipe. A containment vessel for a nuclear reactor. 7. The nuclear power plant according to claim 5, wherein the nuclear power plant is shared with a dry well in each nuclear reactor.
Connect two wet wells or mid wells with a connecting vent pipe, and connect
A reactor containment vessel provided with a check valve or a rupture disk having a pressure support mechanism on the back side. 8. The light water nuclear power plant, wherein two or more reactors are provided in the same facility, and one static safety system is shared by the two or more reactors. Nuclear power facilities. 9. A light water nuclear power plant, wherein two or more reactors are provided in the same facility, and a dryer, separator and pit are shared by the two or more reactors. Power generation equipment. 10. A nuclear power plant having a high-pressure core cooling system, a low-pressure core cooling system, a residual heat removal system, and a reactor auxiliary cooling system as an emergency core cooling system, wherein the high-pressure core cooling system, the low-pressure core cooling A nuclear power plant, wherein the system and the residual heat removal system have a four-section configuration, and the piping of the reactor accessory cooling system has a two-loop configuration. 11. The nuclear power plant according to claim 10, further comprising at least two reactor accessory cooling system pumps and reactor accessory seawater cooling system in the piping of each loop of the reactor accessory cooling system. A nuclear power plant having a configuration and an emergency power supply for supplying power to these components in a four-section configuration. 12. A nuclear power plant wherein the emergency core cooling system according to claim 10 or 11 is shared by two or more nuclear reactors. 13. A reactor comprising both the configuration of the main steam pipe according to claim 1 and the configurations of the reactor pressure vessel and the reactor containment vessel according to claim 4. Nuclear power plant. 14. A plurality of nuclear reactors each having the configuration according to claim 13, wherein the safety system or the configuration of the dryer separator pit according to any one of claims 5 to 9 is combined with these nuclear reactors. Nuclear power facilities characterized by the following.