JP4761988B2 - Boiling water nuclear power plant - Google Patents

Description

  The present invention is capable of cooling the reactor core and the containment vessel for a long period of time in the event of a pipe breakage accident, etc., and in the reactor containment vessel when a reactor emergency shutdown failure event (ATWS event) occurs. The present invention relates to a boiling water nuclear power generation facility capable of preventing a rapid pressure increase.

  In a boiling water nuclear power generation facility equipped with a high pressure resistant reactor containment vessel, as described in Non-Patent Documents 1 and 2, piping connected to the reactor pressure vessel is broken inside the reactor containment vessel. When such an accident occurs, the steam flowing out from the reactor pressure vessel through the fracture port is confined inside the reactor containment vessel, and the reactor pressure vessel and the containment vessel are equalized at a relatively high pressure. Stop the outflow of steam.

  With the steam outflow stopped, the steam generated by the decay heat of the reactor core is led to the emergency condenser, and the steam is stored in advance in the process of passing the steam through the heat transfer pipe of the emergency condenser. Heat exchange with water to condense and return to reactor pressure vessel.

  Then, the reactor core is cooled by maintaining the flooding with the self-holding water previously held as cooling water in the reactor pressure vessel.

Furthermore, in order to maintain the flooding of the reactor core in the long term, the lower part of the reactor containment vessel is flooded with cooling water (self-holding water) that has flowed out of the reactor pressure vessel, and then the reactor pressure vessel is The pressure is reduced and the pressure is equalized with the reactor containment vessel, and the cooling water flowing out to the high pressure reactor containment vessel is returned to the reactor pressure vessel again due to the gravity difference by the piping arranged above the reactor core.
H. Heki, et al., "DEVELOPMENT OF STATU OF CONTAINMENT BWR", Proceedings of ICONE13, 13th International Conference on Nuclear Engineering, Beijing, China, May 15-19, 2005 H. Heki, et al., "DEVELOPMENT OF SIMPLIFIED COMPACT CONTAINMENT BWR PLANT", Proceedings of ICONE12, 12th International Conference on Nuclear Engineering, April 25-29, 2004, Arlington, Viginia USA

  In a boiling water nuclear power plant equipped with a high pressure resistant reactor containment vessel, if the flooding of the reactor containment vessel up to the level of the pressure equalization injection pipe is achieved only with the self-retained water in the reactor pressure vessel, The furnace pressure vessel becomes unrealistically large.

  For this reason, an external tank in which water is stored in advance is installed, the external tank is pressure-equalized with the reactor pressure vessel and the reactor containment vessel, and the water in the external tank is supplied to the reactor pressure vessel and the high pressure reactor. It was necessary to replenish the containment vessel.

  In addition, when non-condensable gas such as nitrogen sealed inside the reactor containment vessel flows into the heat exchanger tube of the emergency condenser through the breakage port and accumulates at the time of a pipe breakage accident, etc. The heat removal performance of the emergency condenser will be significantly degraded.

  In order to avoid the deterioration of the heat removal performance of the emergency condenser, an external tank for gas discharge (vent) is installed outside the reactor containment vessel, and the noncondensable gas from the emergency condenser is installed. It is conceivable to perform venting continuously.

  However, in the unilateral venting from the reactor pressure vessel and the containment vessel to the tank side, once the external tank is equalized with the reactor pressure vessel, the gas cannot be vented any further, and several days after the accident occurred. Long-term cooling that lasts is impossible.

  For this reason, it was necessary to establish a continuous and long-term vent mechanism for non-condensable gas from the emergency condenser.

  Further, when an emergency nuclear reactor emergency stop (scrum) failure event occurs, a so-called ATWS (Anti-Transient Transient Without Scram) event occurs, there is no place to release the steam generated in the reactor core. Therefore, it was necessary to prevent the pressure inside the reactor containment vessel from rapidly increasing and exceeding the allowable pressure.

  The present invention has been made in view of the above problems, and in a boiling water nuclear power generation facility having a high pressure resistant reactor containment vessel, cooling of the reactor core and the reactor containment vessel for a long period of time when a pipe breakage accident occurs, etc. Continuing to have a cooling water replenishment system to the reactor core and reactor containment vessel from the outside in order to achieve the above, and to minimize heat removal performance degradation due to noncondensable gas in the emergency condenser To provide a boiling water nuclear power generation facility having a mechanical and long-term non-condensable gas vent mechanism and means for suppressing a rapid pressure rise inside the containment vessel when an ATWS event occurs Objective.

  In order to solve the above problems, a boiling water nuclear power generation facility according to the present invention includes a reactor containment vessel containing a reactor pressure vessel having a safety valve for overpressure protection, as described in claim 1; A main steam system that sends steam generated in the reactor pressure vessel to a steam turbine, a reactor condensate system that condenses steam from the steam turbine by a condenser, and condensate condensed in the reactor condensate system. In a boiling water nuclear power plant equipped with a reactor water supply system that returns to the reactor pressure vessel, an emergency condensate system that condenses the steam from the reactor pressure vessel and returns it to the reactor pressure vessel again, and An external storage system for storing condensate from an emergency condensate system, the external storage system storing an external tank for storing condensate from the emergency condensate system, and water in the external tank to the reactor Reactor cooling with water injection into the pressure vessel A supplementary water pipe and a reactor containment submersion pipe for injecting water in the external tank into the reactor containment vessel, and the reactor cooling makeup water pipe is connected to the reactor pressure vessel and the external tank. The connected height is set to be different from the height at which the reactor containment submergence piping is connected to the reactor containment vessel and the external tank, and the amount of water injected into the reactor pressure vessel The water injection start timing, the water injection amount to the reactor containment vessel, and the water injection start timing are each driven and controlled by gravity.

  According to the boiling water nuclear power generation facility according to the present invention, cooling of the reactor core and the containment vessel can be maintained for a long time in the event of a pipe breakage accident, and a sudden increase in the inside of the containment vessel when an ATWS event occurs. It is possible to suppress an increase in pressure.

  An embodiment of a boiling water nuclear power generation facility according to the present invention will be described with reference to the accompanying drawings.

[First Embodiment]
FIG. 1 shows a configuration diagram of a first embodiment of a boiling water nuclear power generation facility according to the present invention.

  The boiling water nuclear power generation facility 1 is a small (for example, 300 to 500 MWe class) dry container type boiling water nuclear power generation facility, and as shown in FIG. In addition, the reactor containment vessel 2 accommodates the reactor pressure vessel 3.

  At this time, the reactor containment vessel 2 is supported and fixed by the containment vessel support pedestal 4 which is a pedestal fixed to the floor of the reactor building, and the reactor pressure vessel 3 is contained in the reactor containment vessel 2. It is supported by the pressure vessel support pedestal 5 which is a fixed base.

  In addition, the boiling water nuclear power generation facility 1 includes, for example, two main steam systems 6 that send steam generated in the reactor pressure vessel 3 to a steam turbine (not shown) that generates power, and steam from the steam turbine. A reactor condensate system (not shown) for condensing the water with a condenser (not shown), and a reactor water supply for returning the condensate condensed in the reactor condensate system to the reactor pressure vessel 3 via a feed water pump 7a A system 7 and a safety system 8 that ensures safety when a pipe breakage accident occurs or when an ATWS event occurs are provided.

  The reactor pressure vessel 3 includes a reactor core 9 in the lower part, and the reactor core 9 is surrounded by a core shroud 10. The feed water returned from the reactor feed water system 7 into the reactor pressure vessel 3 becomes reactor water, is guided to the lower core plenum 12 through the downcomer 11, and is inverted by the lower core plenum 12 so that the reactor core 9 is Ascend the street.

  The reactor water is evaporated upon receiving the nuclear reaction heat when it rises in the reactor core 9, becomes a gas-liquid two-phase flow, is flowed above the reactor core 9, and is supplied with steam and water by the steam-water separator 13. Separated.

  The steam separated by the steam separator 13 is further dehumidified by a torus-shaped or sleeve-shaped steam dryer 14 and guided to the main steam system 6 via the main steam isolation valve 15. Further, the water separated by the steam separator 13 flows down the downcomer 11, which is an annular region between the core shroud 10 and the reactor pressure vessel, and is returned to the reactor core 9 again.

  The shape of the reactor containment vessel 2 covers the lower portion of the reactor pressure vessel 3 in a close proximity for the purpose of minimizing the amount of flooding around the reactor pressure vessel 3. A reactor core 9 is provided in the lower part of the reactor pressure vessel 3, and the periphery of the reactor core 9 in the reactor containment vessel 2 is designed in a cylindrical shape. Since pipes and valves from the reactor pressure vessel 3 are arranged in the central portion of the reactor containment vessel 2 in the height direction, the central region of the reactor containment vessel 2 is larger from the reactor pressure vessel 3 in the outer circumferential direction. The bulging sphere that bulges is designed, for example, in a spherical shape, and the upper region of the reactor containment vessel 2 in the height direction is designed in a hemispherical shape.

  A control rod driving device 16 is provided on the upper portion of the reactor pressure vessel 3, and the control rod driving device 16 converts the control rod 18 stored in the control rod storage space 17 above the reactor core 9 into the reactor core. 9 is controlled to be inserted from above.

  By inserting the control rod 18 from the upper part of the reactor core 9, it is possible to insert the control rod 18 into the reactor core 9 quickly using gravity.

  Next, the safety system 8 provided in the boiling water nuclear power generation facility 1 will be described.

  The safety system 8 includes an emergency condensate system 19 that condenses steam from the reactor pressure vessel 3 and returns it to the reactor pressure vessel 3 in an emergency, and an external storage system that stores the condensate from the emergency condensate system 19. 20 and a poison injection system 21 for injecting a poison such as boric acid water into the reactor pressure vessel 3.

  A safety valve 22 is provided above the reactor pressure vessel 3 to release the internal steam into the reactor containment vessel 2 when the pressure inside the reactor pressure vessel 3 becomes abnormally high. Since high-pressure steam is released into the reactor containment vessel 2 through the safety valve 22, the reactor containment vessel 2 is configured to have a high pressure resistance design.

  Moreover, the reactor containment vessel 2 and the reactor pressure vessel 3 are stored in the reactor using the difference in water level (gravity difference) when the reactor containment vessel 2 and the reactor pressure vessel 3 become the same pressure. A communication pipe 23 for supplying cooling water from the vessel 2 to the reactor pressure vessel 3 is provided.

  The reactor pressure vessel 3 is not provided with a nozzle such as a pipe below the upper end of the reactor core 9.

  The emergency condensate system 19 has an emergency condenser pool 24 that absorbs heat, and an emergency condenser 25 that condenses steam is provided inside the emergency condenser pool 24. The emergency condenser 25 includes an inlet water chamber 26 into which steam from the reactor pressure vessel 3 flows, a heat transfer tube that contacts the steam and takes heat from the steam, and an outlet water chamber 27 into which condensate flows from the heat transfer tube. Is provided.

  Further, the emergency condenser 25 contains the emergency condenser steam suction pipe 28 for sucking the steam in the reactor pressure vessel 3 and the cooling water condensed by the emergency condenser 25 to the reactor pressure vessel 3. An emergency condenser condensate return pipe 29 to be injected is installed.

  The emergency condenser 25 discharges the non-condensable gas to prevent the non-condensable gas flowing in with the condensate from the emergency condensate system 19 from accumulating in the heat transfer tube and degrading the heat removal performance. An emergency condenser gas vent pipe 30 and an emergency condenser main vent pipe 31 for discharging the condensate from the emergency condensate system 19 to the external storage system 20 are provided.

  The emergency condenser main vent pipe 31 is set to have a larger diameter than the emergency condenser gas vent pipe 30 in order to prevent the pressure in the reactor containment vessel 2 from rising instantaneously. That is, a plurality of pipes with different diameters are installed in parallel as emergency condenser vent pipes for pouring water from the emergency condenser 25 to the external tank 34.

  The emergency condenser gas vent pipe 30 is provided with an emergency condenser gas vent flow restriction orifice 32 which is a throttle for controlling the flow rate of the non-condensable gas flowing in the pipe. At one end, an emergency condenser main vent pipe sparger 33 is provided. This is because the steam condensing effect is enhanced by providing a sparger structure for the outlet from the emergency condenser main vent pipe 31 to the stored water in the external tank 34.

  Further, the height of the discharge port on the external storage system 20 side of the emergency condenser main vent pipe 31 is set lower than the height of the suction port on the emergency condensate system 19 side.

  The external storage system 20 includes a cooling water stored in advance, condensate flowing out from the emergency condenser main vent pipe 31, and an external tank 34 for storing non-condensable gas. Cooling water is stored in the external tank 34, and the cooling water is used from above the external tank 34 as reactor cooling replenishment water A, reactor containment submerged water B, and vent steam condensation water C.

  The external tank 34 includes a reactor cooling makeup water pipe 35 that injects reactor cooling makeup water A into the reactor pressure vessel 3, and a reactor containment that injects reactor containment submersion water B into the reactor containment vessel 2. And a container submergence pipe 36.

  That is, in the external tank 34, the coolant stored above the reactor cooling makeup water pipe 35 by gravity driving is used as the reactor cooling makeup water A and the reactor containment vessel flooding pipe 36. The cooling water in the upper part is used as the reactor containment vessel submersion water B, and the cooling water in the lower part of the reactor containment vessel submergence pipe 36 is used as the vent steam condensation water C.

  Further, the external tank 34 sucks the cooling water in the external tank 34 through the residual heat removal system pipe 37 and the residual heat that takes away the residual heat from the cooling water sucked up by the residual heat removal system pump 38. There is provided a cooling pipe comprising a removal system heat exchanger 39 and an external tank place purger 40 that disperses the cooling water sucked up by the residual heat removal system pump 38.

  Further, the high pressure makeup water pipe 41 for injecting water from the external tank 34 to the reactor pressure vessel 3 is provided with a high pressure makeup water pump 42, and the outlet portion of the cooling water of the high pressure makeup water pipe 41 is connected to the reactor water system 7. Is done.

  The poison injection system 21 includes a poison injection system accumulating storage tank 43 for storing a poison such as boric acid water to be injected into the reactor pressure vessel 3 in an emergency, and a poison injection system poison for injecting the poison into the reactor pressure vessel 3. And an injection pipe 44.

  Next, Example 1 of the boiling water nuclear power generation facility 1 according to the present invention will be described with reference to FIGS.

  In the first embodiment, an explanation will be given of an event at the time of a coolant loss accident that occurs when an accident occurs in which a pipe or the like connected to the reactor pressure vessel 3 is broken inside the reactor containment vessel 2.

  2 to 4 show the event transition state of the boiling water nuclear power generation facility 1 when the reactor coolant is lost along the time axis. 5 and 6 are graphs showing fluctuations in main parameters such as the water level L1 in the reactor pressure vessel when the reactor coolant is lost.

  In the boiling water nuclear power generation facility 1, steam generated in the reactor pressure vessel 3 normally reaches the steam turbine via the main steam system 6 and is used for power generation. Then, the steam coming out of the steam turbine is condensed by the reactor condensate system to become condensate, and this condensate is returned again to the reactor pressure vessel 3 as feed water by the reactor feed water system 7.

  However, when a rupture accident occurs in a pipe connected to the reactor pressure vessel 3, steam flows out from the reactor pressure vessel 3 to the reactor containment vessel 2 through the rupture port 45, and this steam is atomized. It is confined inside the furnace containment vessel 2.

  At this time, as shown in FIG. 2, while the water level in the reactor pressure vessel 3 falls, the pressure and temperature in the reactor containment vessel 2 and the temperature of the vessel wall of the reactor containment vessel 2 rise. Further, a part of the steam that has flowed into the reactor containment vessel 2 is condensed and accumulated at the bottom of the reactor containment vessel 2, and the coolant level in the reactor containment vessel 2 rises.

  As a result of the steam in the reactor pressure vessel 3 flowing out into the reactor containment vessel 2, the pressure in the reactor pressure vessel 3 and the reactor containment vessel 2 is quickly equalized, and the outflow of steam from the fracture port 45 is stopped. To do.

  In a state where the steam is confined in the reactor containment vessel 2, the steam generated by the decay heat of the reactor core 9 is led to the emergency condenser 25 via the emergency condenser steam suction pipe 28, and the emergency condensate In the process of passing steam through the heat transfer pipe of the condenser 25, the steam is condensed by exchanging heat with the water stored in the emergency condenser pool 24, and from the emergency condenser condensed water return pipe 29 to the reactor pressure vessel 3. return.

  As described above, immediately after the accident occurs, that is, relatively early after the accident occurs, the reactor core 9 is cooled by reducing the pressure in the reactor containment vessel 2 and the reactor pressure vessel 3, and the reactor pressure before the accident occurs. This is achieved by maintaining submergence with self-retained water retained as cooling water inside the container 3.

  At this time, since the heat removal capability of the emergency condenser 25 is superior to the amount of heat released from the reactor containment vessel 2, the pressure drop in the reactor pressure vessel 3 is lower than the pressure drop in the reactor containment vessel 2. As soon as possible, the nitrogen gas enclosed in the reactor containment vessel 2 before the accident occurs flows into the reactor pressure vessel 3 through the fracture opening 45.

  The nitrogen gas is sucked into the emergency condenser 25 through the emergency condenser steam suction pipe 28 together with the steam, and discharged (vented) to the external tank 34 through the emergency condenser gas vent pipe 30. Along with the flow of condensate condensed by the emergency condenser 25, it is recirculated to the reactor pressure vessel 3 (state (I) in FIG. 5).

  On the other hand, when the mixed fluid of non-condensable gas such as steam and nitrogen gas flowing into the emergency condenser 25 flows through the heat transfer pipe of the emergency condenser 25, the vapor of the mixed gas is condensed. Since the partial pressure of the non-condensable gas increases, the heat removal performance of the emergency condenser 25 deteriorates as it goes downstream of the heat transfer tube.

  The partial pressure of the non-condensable gas is maximized at the outlet of the heat transfer tube of the emergency condenser 25 and is discharged into the outlet water chamber 27 together with the remaining steam and condensed water. By selectively venting the mixed fluid having a high non-condensable gas partial pressure to the external tank 34, the heating of the stored water in the external tank 34 is minimized, and the gas phase pressure is gradually increased. The reactor containment vessel 2 and the reactor pressure vessel 3 are finally equalized over several hours to about a day after the accident.

  When a long time has passed since the occurrence of the accident, as shown in FIG. 3, the water level L <b> 1 in the reactor pressure vessel 3 gradually decreases due to the outflow of steam and approaches the upper end of the reactor core 9. In order to recover the water level L1 in the reactor pressure vessel 3, the reactor pressure vessel 3 is poured into the reactor pressure vessel 3 by a gravity difference from the external pressure tank 3 and the external pressure tank 34 through the reactor cooling make-up water pipe 35. (State (II) in FIG. 5).

  The suction of the reactor cooling makeup water pipe 35 opens to a predetermined height of the external tank 34 so that a necessary amount can be replenished. In some cases, water is injected into the reactor pressure vessel 3. There are times when it is not. FIG. 6 is a graph showing fluctuations in main parameters when the reactor coolant is lost when water injection from the external tank 34 to the reactor core 9 is not performed.

  If a further long time elapses after water injection from the external tank 34 to the reactor pressure vessel 3, heat radiation to the outside of the reactor containment vessel 2 continues, so if left as it is, the water level L1 in the reactor pressure vessel 3 is There is a possibility that the reactor core 9 will be exposed unless it is lowered gradually and some measures are taken.

  Therefore, in order to shift the reactor core 9 to a stable cooling state, as shown in FIG. 4, the reactor containment vessel 2 is moved from the external tank 34 through the reactor containment vessel submerged piping 36 by the difference in gravity. The container 3 and the emergency condenser condensate return pipe 29 are submerged above the connection height.

  The water level L1 in the reactor pressure vessel is maintained above the reactor core 9 by replenishing the reactor pressure vessel 3 from the reactor containment vessel 2 through the communication pipe 23 to the reactor pressure vessel 3 by gravity difference.

  The suction of the reactor containment submerged pipe 36 opens to a predetermined height below the reactor cooling supply water pipe 35 of the external tank 34 so that a necessary amount can be supplied.

  Once the cooling circulation path through the communication pipe 23 is formed, water from other than these systems can be used as long as the predetermined heat removal is performed by the steady discharge of the accumulated gas in the emergency condenser 25. The reactor core 9 can be continuously cooled without replenishing (state (III) in FIG. 5).

  Further, since the communication pipe 23 is arranged outside the reactor containment vessel 2, it can be designed without considering the atmospheric conditions in the reactor containment vessel 2 that become high temperature or high pressure in the event of an accident, and has high reliability. This opening operation is realized with

  Returning the non-condensable gas staying in the emergency condenser 25 to the reactor pressure vessel 3 along with the condensed water results in the piping sizing of the emergency condenser condensate return pipe 29 and the reactor pressure vessel 3. It becomes possible by appropriately setting the height of the U-seal (U-shaped pipe) in front of the connecting part.

  In general, the higher the condensate flow rate, the easier the accompanying non-condensable gas will be discharged, while the faster the water and gas flow rate, the greater the pressure loss in the emergency condenser condensate return pipe 29. Becomes larger.

  The flow rate of the steam flowing into the emergency condenser 25, that is, the heat removal performance of the emergency condenser 25 is determined by the connection end of the emergency condenser steam suction pipe 28 and the reactor pressure vessel 3, and the emergency condenser. The connection end of the condensed water return pipe 29 and the reactor pressure vessel 3, the head difference between the reactor pressure vessel 3 side and the emergency condenser 25 side between the two connection ends, and the system of the emergency condenser 25 Since it is determined by the pressure loss of the pipe, an increase in the pressure loss of the emergency condenser condensate return pipe 29 may hinder the heat removal performance of the emergency condenser 25.

  Therefore, in order to efficiently discharge the noncondensable gas from the emergency condenser 25 along with the condensed water at the time of an accident without hindering the heat removal of the emergency condenser 25, the emergency condenser is used. It is necessary to select the diameter of the condenser condensate return pipe 29 appropriately.

  When the emergency condenser condensate return pipe 29 is filled with condensate and flows under conditions where the emergency condenser 25 satisfies a predetermined heat removal performance under a high pressure condition of 7 MPa near the reactor rated operating pressure, At the diameter of the condensate return pipe 29 for emergency condensers where the average flow velocity of the condensed water is equal to or higher than a predetermined value, for example, 1 meter per second, non-condensable gas is efficiently discharged while maintaining heat removal performance. Can satisfy one request.

  Further, it is necessary to prevent the backflow of steam from the connection portion of the emergency condenser condensate return pipe 29 and the reactor pressure vessel 3 toward the emergency condenser 25. The water return pipe 29 is connected to the reactor pressure vessel 3 as a U-shaped pipe (U seal). On the other hand, the height of the U seal is from the viewpoint of discharging non-condensable gas from the emergency condenser 25. Affects.

  From this, by making the height of the U seal coincide with the pipe diameter (1D), the exhaust of the non-condensable gas from the emergency condenser 25 can be promoted while maintaining the effect of the U seal. .

  Even if the reactor core 9 is damaged by this function and the reactor core is damaged, zirconium and the fuel cladding tube material react with steam to generate a large amount of hydrogen. The hydrogen gas flowing into the reactor 25 is continuously discharged to the reactor pressure vessel 3 without accumulating in the heat transfer tube.

  Therefore, according to the first embodiment, the heat removal performance of the emergency condenser 25 is not extremely deteriorated, and continuous heat removal from the reactor containment vessel 2 is possible.

  Next, a second embodiment of the boiling water nuclear power generation facility 1 according to the present invention will be described with reference to FIGS.

  In the second embodiment, an event when an event (hereinafter referred to as an ATWS event) that has failed in an emergency reactor shutdown (scrum) has occurred will be described.

  7 to 9 show the event transition along the time axis when an ATWS event occurs, taking as an example the case where the main steam isolation valve is closed. FIG. 10 is a graph showing fluctuations in main parameters during ATWS.

  When the main steam isolation valve is closed for some reason during the operation of the reactor, an interlock that automatically scrams the reactor is usually provided. However, if this scram fails, the pressure in the reactor pressure vessel 3 rises due to the steam generated from the reactor core 9, and further, the steam is ejected from the safety valve 22, causing the pressure and temperature of the reactor containment vessel 2 to suddenly increase. It will rise.

  In order to cope with the pressure rise in the reactor containment vessel 2 and the reactor pressure vessel 3, conventionally, means (such as a pressure suppression method) such as restricting the feed water flow rate to the reactor pressure vessel 3 to the maximum using a pressure suppression pool are used. It was used to reduce the reactor output by reducing the water level in the reactor pressure vessel 3.

  However, since the boiling water nuclear power generation facility 1 employs a dry container method and does not have a pressure suppression pool, the number of events after the occurrence of an event in which the pressure in the reactor containment vessel 2 does not sufficiently lower the reactor output L6. There is a possibility of increasing to the pressure of the reactor pressure vessel 3 in minutes.

  For this reason, in the boiling water nuclear power generation facility 1, a large-diameter emergency condenser main vent pipe 31 is provided in the outlet water chamber of the emergency condenser 25, and steam generated from the reactor core 9 is externally supplied. It was set as the structure which guides in the storage water of the tank 34 and condenses.

  By operating this vent almost simultaneously with the detection of the occurrence of the ATWS event and reducing the feed water flow rate, in addition to the heat removal by the emergency condenser 25, the removal by steam condensation in the external tank 34 is performed. Heat can also be implemented.

  As shown in FIG. 10, when the ATWS event occurs, the pressure L2 in the reactor pressure vessel 3 and the pressure L3 in the reactor containment vessel 6 do not increase excessively. Therefore, in order to maintain this pressure at the minimum, Reducing the output L6 of the reactor by lowering the water level L1 in the reactor pressure vessel 3 (state (II) in FIG. 10).

  In the period up to this point, an attempt is made to insert the control rod 18 using a scram or other means by the operator's operation. If all of these fail, the process proceeds to the next stage.

  Further, as shown in FIG. 8, the high-pressure makeup water pump 42 prevents the pressure from rising excessively as the water level of the external tank 34 rises and the volume of the gas phase portion decreases during the period so far. Etc. to adjust the external tank water level (state (III) in FIG. 10).

  Furthermore, by causing the residual heat removal system pump 38, the residual heat removal system heat exchanger 39, and the external tank space purger 40 provided in the external tank 34 to function, an increase in the water temperature of the external tank 34 due to steam condensation is suppressed.

  After the above time has elapsed, poison is rapidly injected into the region of the reactor core 9 from the poison pressure storage tank 24 as the final reactor shutdown means. As a result, the nuclear reactor can be promptly shifted to a high temperature subcritical state (state (IV) in FIG. 10).

  By appropriately setting the poison injection time, the emergency condenser vent capacity, the external tank capacity, and the like, the water temperature and pressure in the external tank can be kept below a predetermined value even during the period up to this stage.

  After the reactor core 9 becomes a high-temperature subcritical state by poisoning, it is possible to shift to low-temperature subcriticality by injecting further poison.

  Further, as shown in FIG. 9, after reaching high temperature subcriticality, the water level L1 in the reactor pressure vessel 3 is gradually recovered by water injection from the external tank 34 while heat removal by the emergency condenser 25 is performed. By doing so, a safe state can be maintained (state (V) in FIG. 10).

  Even when an event occurs in which all the functions of the emergency condenser 25 are lost, the external tank 34 is simultaneously depressurized by venting the steam generated from the reactor core 9 to depressurize the reactor pressure vessel 3. Water is sucked by the residual heat removal system pump 38, cooled by the residual heat removal system heat exchanger 39, and then returned by the external tank place purge 40. Further, by combining the maintenance of the water level in the reactor pressure vessel 3 with the high-pressure make-up water pump 42, the reactor pressure vessel 3 can be safely reduced to a pressure at which the operation for cooling when the residual heat removal system pump 38 is stopped can be performed. It becomes possible to do.

[Second Embodiment]
Next, a second embodiment of the boiling water nuclear power generation facility according to the present invention will be described with reference to FIG. In addition, the same code | symbol is attached | subjected to the structure same as 1st Embodiment, and detailed description is abbreviate | omitted.

  The boiling water nuclear power generation facility 1A of the second embodiment has added a back-up function for suppressing the pressure increase of the reactor containment vessel 2 during ATWS in the boiling water nuclear power generation facility 1 of the first embodiment. Is.

  That is, as shown in FIG. 11, the safety system 8A of the boiling water nuclear power generation equipment 1A is arranged outside the reactor containment vessel 2 in order to suppress the pressure rise in the reactor containment vessel 2 and the external tank 34. A water pool 46 different from the external tank 34 provided is provided.

  In addition, a high pressure-resistant nuclear reactor containment safety valve 47 is installed in the reactor containment vessel 2, and the reactor containment vessel safety valve 47 supplies exhaust gas from the reactor containment vessel 2 to the liquid phase portion of the water pool 46. An exhaust pipe 54 leading to 46a is provided. A high pressure resistant reactor containment vessel safety valve exhaust pipe sparger 48 is provided at the opening of the exhaust pipe 54 on the water pool 46 side.

  Further, an external tank safety valve 49 is also installed in the gas phase portion of the external tank 34. Similarly, the exhaust from the external tank safety valve 49 is discharged from the exhaust pipe 55 to the liquid phase section 46a of the water pool 46. Through 50.

  In the boiling water nuclear power generation facility 1A of the second embodiment, when steam discharge from the emergency condenser main vent pipe 31 to the external tank 34 fails during ATWS, the reactor containment vessel 2 is connected via the safety valve 22. The discharged reactor steam is led from the reactor containment vessel safety valve 47 to the water pool 46 and condensed, and the nitrogen gas inside the reactor containment vessel 2 discharged together with this steam can also be stored in the water pool 46. Thereby, the excessive pressure rise of the reactor containment vessel 2 can be suppressed.

  Even when the steam is successfully discharged from the emergency condenser main vent pipe 31 to the external tank 34, the residual heat removal system pump 38, the residual heat removal system heat exchanger 39 and the outside disposed in the external tank 34. If cooling of the water inside the external tank 34 using the tank displacement purger 40 fails, the pressure in the external tank 34 increases. However, when the pressure in the external tank 34 reaches the operating pressure of the external tank safety valve 49, By transferring non-condensable gas such as steam and nitrogen gas to the water pool 46, the pressure increase in the external tank 34 can be suppressed, and the pressure increase in the reactor pressure vessel 3 by the emergency condenser main vent pipe 31 Suppression is also possible.

  As a result, an excessive pressure rise in the reactor containment vessel 2 can be suppressed in a period until the reactor is stopped by the poison (boric acid water or the like) injected from the poison injection system pressure accumulation storage tank 43, After the reactor is shut down, the reactor core 9 can be cooled by condensing steam generated by the decay heat of the reactor core 9 by the emergency condenser 25.

  Further, such a method for suppressing the pressure increase of the containment vessel 2 is not only during ATWS, but also causes core damage due to insufficient cooling of the reactor core 9, resulting in a large amount of reaction between zirconium, which is a fuel cladding material, and steam. It can be used even when hydrogen is generated.

  In addition, the water pool 46 can utilize the condensate storage tank used when exchanging the fuel in the reactor pressure vessel 3, for example. In this case, in order to perform the method for suppressing the pressure increase of the containment vessel 2 of the first embodiment, the containment vessel safety valve 47, the external tank safety valve 49, the reactor containment vessel safety valve exhaust pipe sparger 48, the external tank safety valve exhaust. By simply adding the sparger 50 for pipes and the exhaust pipes 54 and 55 associated therewith, it is possible to obtain a back-up function for suppressing pressure rise during ATWS, and to further increase safety.

[Third Embodiment]
Next, a third embodiment of the boiling water nuclear power generation facility according to the present invention will be described with reference to FIG. In addition, the same code | symbol is attached | subjected to the structure same as 1st Embodiment and 2nd Embodiment, and detailed description is abbreviate | omitted.

  In the boiling water nuclear power generation facility 1B of the third embodiment, assuming that the functions of all the emergency condensers 25 are lost for some reason in the boiling water nuclear power generation facility 1 of the first embodiment, By providing an alternative cooling means as a preparation after that, safety is further improved.

  That is, the safety system 8B of the boiling water nuclear power generation facility 1B includes a water pool 46 different from the external tank 34 outside the reactor containment vessel 2 as shown in FIG.

  The safety system 8B includes a safety valve 51 with an automatic pressure reducing function and a safety valve with an automatic pressure reducing function for discharging steam generated from the reactor core 9 to the external tank 34 when the functions of all the emergency condensers 25 are lost. A safety valve exhaust pipe 52 with an automatic pressure reducing function for guiding exhaust from 51, and a safety valve exhaust pipe sparger 53 with an automatic pressure reducing function for discharging the exhaust gas guided to the safety valve exhaust pipe 52 with an automatic pressure reducing function dispersed in the external tank 34. Prepare.

  The steam generated from the reactor core 9 is discharged to the external tank 34 through the safety valve 51 with an automatic pressure reducing function, and the reactor pressure vessel 3 is removed from heat.

  As a result, the outflow of cooling water or steam from the safety valve 22 or the breakage opening 45 to the reactor containment vessel 2 is suppressed to the minimum, and the rise in pressure and temperature of the reactor containment vessel 2 and the reactor pressure vessel 3 is suppressed. The

  At the same time, in order to keep the reactor core 9 submerged, the water in the external tank 34 is injected from the reactor water supply system 7 to the reactor pressure vessel 3 using the high-pressure make-up water pump 42, or the water in the external tank 34 is Is achieved by injecting water into the reactor containment vessel 2 from the reactor containment vessel submerging pipe 36 and opening the reactor containment vessel communication pipe 23B into the reactor pressure vessel 3.

  The water in the external tank 34 is cooled by the residual heat removal system pump 38 and the residual heat removal system heat exchanger 39, and the reactor pressure vessel 3 and the reactor containment vessel 2 are cooled by cooling the water in the external tank 34. Done.

  Further, by providing this alternative cooling means, it is not necessary to provide the emergency condenser main vent pipe 31.

  Note that the boiling water nuclear power generation facility 1B of the third embodiment can also be used as a cooling means for the reactor pressure vessel 3 and the reactor containment vessel 2 during ATWS.

  First, when the reactor is isolated due to the failure of the scram, the pressure in the reactor pressure vessel 3 rises and the safety valve function of the safety valve 51 with the automatic decompression function works to automatically transfer the steam in the reactor pressure vessel 3 to the external tank 34. To exhaust.

  When the capacity is insufficient with this discharge function alone, the safety valve 22 installed in the reactor pressure vessel 3 is further opened, the steam in the reactor pressure vessel 3 is discharged into the reactor containment vessel 2, and the pressure is reduced. Raise. At this time, a safety valve 22 having a sufficient capacity is installed so that the pressure in the reactor containment vessel 2 is equal to or lower than the design pressure.

  After the safety valve 22 is opened, the reactor power drop caused by the water level drop in the reactor pressure vessel 3 due to the discharge of the steam in the reactor pressure vessel 3, and the poison after the steam discharge in the reactor pressure vessel 3 The reactor is stopped by the injection, and is cooled to the high temperature standby state by the cooling by the emergency condenser 25, and the event is converged.

BRIEF DESCRIPTION OF THE DRAWINGS The block diagram which shows 1st Embodiment of the boiling water nuclear power generation equipment which concerns on this invention. The 1st figure which shows the state of the event water level at the time of the reactor coolant loss of 1st Embodiment. The 2nd figure which shows the state of the event water level at the time of the reactor coolant loss of 1st Embodiment. The 3rd figure which shows the state of the event water level at the time of the reactor coolant loss of 1st Embodiment. The graph which shows the fluctuation | variation of the main parameters at the time of the reactor coolant loss in 1st Embodiment. The graph which shows the fluctuation | variation of the main parameters at the time of the reactor coolant loss in 1st Embodiment. The 1st figure which shows the event transition at the time of ATWS in 1st Embodiment. The 2nd figure which shows the event transition at the time of ATWS in 1st Embodiment. The 3rd figure which shows the event transition at the time of ATWS in 1st Embodiment. The graph which shows the fluctuation | variation of the main parameters at the time of ATWS in 1st Embodiment. The block diagram which shows 2nd Embodiment of the boiling water type nuclear power generation equipment which concerns on this invention. The block diagram which shows 3rd Embodiment of the boiling water nuclear power generation equipment which concerns on this invention.

Explanation of symbols

1, 1A, 1B Boiling water nuclear power plant 2 Reactor containment vessel 3 Reactor pressure vessel 4 Containment vessel support pedestal 5 Pressure vessel support pedestal 6 Main steam system 7 Reactor feed water system 7a Feed water pump 8, 8A, 8B Safety system System 9 Reactor core 10 Core shroud 11 Downcomer 12 Lower core plenum 13 Steam / water separator 14 Steam dryer 15 Main steam isolation valve 16 Control rod drive 17 Control rod storage space 18 Control rod 19 Emergency condensate system 20 External storage System 21 Poison injection system 22 Safety valve 23, 23B Communication piping 24 Emergency condenser pool 25 Emergency condenser 26 Inlet water chamber 27 Outlet water chamber 28 Emergency condenser steam suction piping 29 Emergency condenser condensate Return pipe 30 Emergency condenser gas vent pipe 31 Emergency condenser main vent pipe 32 Emergency condenser gas vent flow restriction orifice 3 Emergency condenser main vent piping sparger 34 External tank 35 Reactor cooling makeup water piping 36 Reactor containment submerged piping 37 Residual heat removal system piping 38 Residual heat removal system pump 39 Residual heat removal system heat exchanger 40 External tank displacement purger 41 High pressure makeup water pipe 42 High pressure makeup water pump 43 Poison injection system pressure storage tank 44 Poison injection system poison injection pipe 45 Break port 46 Water pool 46a Liquid phase part 47 Primary containment vessel safety valve 48 Reactor containment vessel Safety valve exhaust pipe sparger 49 External tank safety valve 50 External tank safety valve exhaust pipe sparger 51 Safety valve with automatic decompression function 52 Safety valve exhaust pipe with automatic decompression function 53 Safety valve exhaust pipe sparger with automatic decompression function 54, 55 Exhaust pipe A Replenishment water for reactor cooling B Reactor containment submersion water C Vent steam condensation water L1 Water level in the reactor pressure vessel L2 Pressure in the reactor pressure vessel L3 Pressure in the reactor containment vessel L4 Pressure in the external tank L5 Water level in the reactor containment vessel L6 Reactor output L7 Water temperature in the external tank

Claims (13)

A reactor containment vessel containing a reactor pressure vessel having a safety valve for overpressure protection, a main steam system for sending steam generated in the reactor pressure vessel to a steam turbine, and a steam from the steam turbine being a condenser In a boiling water nuclear power generation facility comprising a reactor condensate system that condenses by the reactor and a reactor water supply system that returns the condensate condensed in the reactor condensate system to the reactor pressure vessel,
An emergency condensate system that condenses the steam from the reactor pressure vessel and returns it back to the reactor pressure vessel;
An external storage system for storing condensate from the emergency condensate system,
The external storage system includes an external tank for storing condensate from the emergency condensate system, a reactor cooling makeup water pipe for injecting water in the external tank into the reactor pressure vessel, and an internal tank in the external tank. A reactor containment submerged pipe for injecting water into the reactor containment vessel,
The height at which the reactor cooling makeup water pipe is connected to the reactor pressure vessel and the external tank is higher than the height at which the reactor containment submersion piping is connected to the reactor containment vessel and the external tank. The water injection amount and the water injection start timing to the reactor pressure vessel, and the water injection amount and the water injection start timing to the reactor containment vessel are respectively driven and controlled by gravity. Boiling water nuclear power generation facility. The boiling water nuclear power plant according to claim 1, further comprising a poison injection system for injecting a previously stored poison into the reactor pressure vessel in an emergency. The reactor pressure vessel includes a reactor core and an upper insertion type control rod driving device, and only above the upper end of the reactor core, the reactor cooling makeup water pipe, the reactor water supply system pipe, and The boiling water nuclear power plant according to claim 1, wherein the emergency condensate piping is connected to the reactor pressure vessel. An emergency condenser vent pipe for injecting the condensate of the emergency condensate system into the external tank;
The boiling water nuclear power plant according to claim 1, wherein the emergency condenser vent pipe is constructed by arranging a large-diameter pipe and a small-diameter pipe in parallel. The outlet to the outside tank of the emergency condenser vent pipe has a sparger structure, and the height of the outlet is the suction outlet on the emergency condenser side of the emergency condenser vent pipe. The boiling water nuclear power generation facility according to claim 4, which is set below the height. A communication pipe for pouring the submersion in the reactor containment vessel into the reactor pressure vessel;
This communication pipe is connected to an emergency condenser condensate return pipe for injecting condensate from the reactor containment vessel and the emergency condensate system into the reactor pressure vessel, and to the reactor pressure vessel . The boiling water nuclear power plant according to claim 1, wherein the water injection amount and the water injection start timing are driven and controlled by gravity. 2. The boiling water nuclear power generation facility according to claim 1, wherein the external tank includes a cooling pipe including a pump, a pipe, and a heat exchanger. The diameter of the emergency condensate is set so that the average condensate flow velocity in the condensate return pipe of the emergency condenser for injecting the condensate from the emergency condensate into the reactor pressure vessel is 1 meter per second or more. The boiling water nuclear power generation facility according to claim 1, further comprising a condenser. The boiling water according to claim 8, wherein a U-seal having a height approximately equal to the diameter of the emergency condenser condensate return pipe is provided at a connection location of the emergency condenser condensate return pipe to the reactor pressure vessel. Type nuclear power plant. The reactor containment vessel includes a water pool different from the external tank outside the reactor containment vessel, and the reactor containment vessel includes a reactor containment vessel safety valve whose exhaust pipe is connected to a liquid phase part of the water pool, The boiling water nuclear power generation facility according to claim 1, wherein a tip of a connection portion of the exhaust pipe with respect to the water pool has a sparger structure. 11. The external tank safety valve for connecting an exhaust pipe to the liquid phase part of the water pool is provided in the gas phase part of the external tank, and the tip of the connection part of the exhaust pipe to the water pool has a sparger structure. Boiling water nuclear power plant. The reactor pressure vessel is provided with a safety valve with an automatic pressure reducing function for connecting an exhaust pipe to the liquid phase part of the external tank, and when the function of the emergency condensate system is lost, steam generated from the reactor core is The boiling water nuclear power generation facility according to claim 1, wherein the water is discharged to the external tank through an exhaust pipe. The boiling water according to claim 12, wherein a communication pipe for injecting water submerged in the reactor containment vessel into the reactor pressure vessel by gravity difference is installed above the reactor core of the reactor pressure vessel. Type nuclear power plant.

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