JPS58143297A - Feedwater device for bwr type reactor - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION [Technical outline of the invention] The present invention relates to a water supply system for a boiling water nuclear power plant, and particularly relates to a water supply system that can improve fuel cooling performance in the event of a loss of coolant accident. .

[Technical background of the invention]

Figure 1 shows the system and diagram of a water supply system for a conventional boiling water nuclear power plant, in which steam generated in the reactor 1 is transferred to the main steam pipe 2.
The steam is led to the turbine 4 via the main steam isolation valve 8, where it performs work and drives the generator 5.The steam that has performed work in the turbine 4 is condensed in the condenser 6 and stored in the hot well 7. It will be done. The condensate stored in the hot well 7 is drawn out by a condensate pump 8, and is passed through a valve 9, a condensate desalination device 10. Measure 11, [pass through 2, high pressure 1 water pump 13
The water is then pressurized by a valve 14.15 and fed to a low-pressure feedwater heater 16. The water supplied to the low-pressure feedwater heater 16 is heated there to a predetermined temperature, passes through a valve 17, and is sucked into a turbine-driven VJJ water pump 19 driven by a water supply turbine 18 operated by main steam.
There, it is pressurized, passes through the check valve 20 and valves 21 and 22, is further heated by the high-pressure feed water heater 23, and is returned to the reactor 1 through the check valve 20 and the check valve 25. In this way, during normal operation, a feedwater flow rate approximately equal to the main steam flow rate is maintained.

Therefore, if the turbine-driven water supply pump 19 stops operating due to some reason in such a state, a motor-driven water supply pump that is installed in parallel with the turbine-driven water supply pump 19 and the check valve 20 and driven by the electric motor 26 is activated. 27 is activated by the "Feed Water Flow Low" signal, and the feed water heated by the low pressure feed water heater 16 is activated by the above-mentioned motor-driven feed water pump 27 and check valve 28ft.
In addition, the reactor 1 is provided with a water level gauge 29 and a pressure gauge 30, and the detection signals from the water level gauges 29 and 30 are used to control the feed water turbine 18. In order to control the water level and pressure of the coolant inside the reactor to a specified level, and to perform forced convection of the coolant inside the shunted nuclear reactor and the reactor, the cooling bong 31 is included. A recirculation system 32 is provided.

By the way, if the pipe connected to the shoji-l were to break, the coolant inside the reactor would be lost, resulting in a so-called coolant loss accident.

The most severe type of coolant loss accident is a recirculation system piping rupture accident.When an instantaneous guillotine rupture occurs in this piping, the flow rate of the coolant flowing out is large, so it takes only a few tens of seconds for the rupture to occur. fuel exposed.

This causes an increase in the temperature of the fuel and cladding. Therefore, in order to prevent the abnormal temperature rise of the fuel and cladding tube, an emergency core cooling system (not shown) is provided, and in the event of a loss of coolant accident, the emergency core cooling system is activated and the reactor is Coolant (!--pressure injection) is used to cool the core and keep the temperature of the fuel cladding etc. below a predetermined temperature.

[Problems with background technology]

As mentioned above, in the event of a loss of coolant accident, sufficient cooling of the reactor core can be achieved by operating only the emergency core cooling system mentioned above, and water injection into the reactor through the water supply system is not expected. However, even in a situation where a loss of coolant accident occurs and the reactor water level drops, there is no electrical circuit installed that would stop the water supply system, and unless there is a failure in the equipment, the water supply system will continue to operate. The water supply to the reactor can be read. In other words, after a loss of coolant accident occurs, the reactor water level decreases,
When the water level reaches this level, the reactor is forced to scram.

No response is stopped. Therefore, when the water level further decreases and reaches this level, the main steam isolation valve 8 is closed. Then, the rotational speed of the turbine-driven water supply pump 19 decreases by 5 as the pressure of the driving steam decreases, and the electric motor 26 is activated by a signal indicating that the water supply flow rate is low based on this.
The motor-driven water supply pump 27 is activated, thereby continuing water supply. However, when water is supplied by the wake-up machine and the drive water supply pump 27, the water supply flow rate is about 50% of the rated value.

By the way, the feed water temperature is about 22 Or during normal operation, but after the main steam isolation valve 8 is closed, the feed water temperature gradually decreases because a part of the main steam is used as a heat source for the feed water heater. . On the other hand, the reactor pressure is approximately ro during normal operation.
f/d9, but the pressure will be reduced after a loss of coolant accident occurs. Therefore, in the pressure TOVq/1yd9 during normal operation, the feed water temperature 220C is an unsaturated temperature, so the feed water is supplied to the reactor in a completely liquid state, but when the pressure rises, the pressure changes as described above. Since both feed water temperatures decrease, depending on the rate of decrease, the feed water changes from unsaturated water to saturated water to a two-phase mixed fluid to the reactor.

In particular, in the event of a major break in the recirculation piping, which has a severe impact on the reactor fuel, the reactor depressurizes quickly, and during the depressurization, the feed water is depressurized to 191iIIIl, leaving the reactor with saturated water and saturated steam. As a result, the rate of decrease in reactor pressure becomes slower. Therefore, due to the relationship between pump flow rate and discharge pressure, a sufficient flow rate of cold water from the emergency core cooling system into the reactor cannot be obtained, making it impossible to completely cool down the core fuel immediately after a pipe rupture. There are problems such as in some cases.

[Purpose of the invention]

In view of these points, the present invention completely improves the water injection and cooling performance of the water supply system in the event of a loss of coolant accident at a boiling water nuclear power plant, and cools the core fuel by cooperating with the emergency core cooling system. The purpose of the present invention is to provide a water supply system for a nuclear power plant with significantly improved performance and safety of a nuclear reactor.

[Summary of the invention]

The present invention comprises a bypass line @1 that bypasses a high-pressure feedwater heater installed in a water supply system to a nuclear reactor, a bypass line @2 that bypasses a low-pressure feedwater heater and a feedwater pump, and It has an eighth bypass line that bypasses the condensate pump, and a control device that is activated by water level and pressure signals in the reactor and opens each of the bypass lines, and in response to a decrease in the water level, etc. It is characterized in that each bypass pipe is opened in sequence and flood control water is supplied to the reactor through the bypass pipe.

[Example of invention]

FIG. 2 is a schematic system diagram showing an embodiment of the present invention,
Condensate water condensed in the condenser 6 and stored in the hot well 7, a water removal device 10, a high pressure condensate pump 13, a low pressure feed water heater 16, a turbine driven water pump 19, and a prime mover driven water pump 27 The water is supplied to the reactor 1 through a parallel circuit with the high pressure feed water heater 23 and the like. These configurations are exactly the same as the conventional device shown in FIG. 1, but in the present invention, 4P33 and A first valve having a check valve 34
A bypass conduit 35 is provided.

In addition, a valve 15 provided on the inlet side of the low pressure feed water heater 16
A low pressure feed water heater 1 is connected between the upstream side of the feed water heater 1 and the downstream side of the valve 21 provided on the outlet side of the parallel circuit of the feed water pump 19.27.
6. Water supply pump 19. '27 parallel circuit, and a second bypass line 3 that bypasses the above 115 and '21
A check valve 37, a check valve 38, and a check valve 39 are provided in the second bypass conduit 36. Furthermore, the upstream side of the valve 12 on the inlet side of the high-pressure water supply pump 13 and the intermediate portion between the check valve 37 and the valve 38 provided in the second bypass pipe line 36 include a valve 40 and a check valve 41. , are connected by an eighth bypass line 42 that bypasses the high-pressure water supply pump 13 and the valves 12.14 before and after the high-pressure water pump 13.

On the other hand, the production signal from the water level gauge 29 provided in the reactor 1 is applied to the first control device 43, and from the first control device 1ia3, when the water level has decreased to a predetermined value, the high-pressure feed water heating A control signal is output to close the valves 22 and 24 provided before and after the device 230 and open the valve 33 of the first bypass pipe wr35. Further, the pressure detection signal from the pressure gauge 30 is applied to the second control device 44. The second control device 44 is configured to also receive a detection signal from the water level gauge 29, and the second control device 44 is configured to receive a detection signal from the water level gauge 29 when the detection signal from the water level gauge 29 is below a predetermined value. At the same time, when the pressure detection signal from the pressure gauge 30 becomes a signal of, for example, 40s/r*l or less, the valves 15 and 21 of the water supply system are closed and the valves 15 and 21 of the water supply system are closed.
A control signal for opening the valve 38 of the second bypass pipe 36 is output. Further, the second control device 44 detects that the ejection signal from the water level gauge 29 is below a predetermined value, and the pressure detection signal is a lower value, for example, 7 kg/c1.
When the signal becomes less than /lf, the configuration is such that a control signal is output to close the valves 12.14 before and after the high-pressure condensate pump 13 and open the valve 40 of the eighth bypass pipe 42.

In the unlikely event that a reactor loss of coolant accident occurs and the reactor water level drops, the reactor is forced to scram, and when the water level further drops by 6i and reaches a predetermined value, the main steam isolation valve 8
is closed, the first control device t43 closes the valve 22.24 and opens the valve 33 of the first bypass line 35. Therefore, the feed water sent out by the turret-driven water supply pump 27 bypasses the high-pressure feed water heater 23 and is supplied to the reactor via the first bypass pipe 35 and the check valve 25. Therefore, the feed water that can be supplied to the nuclear reactor is not heated by the high-pressure feed water heater 23, its temperature decreases, and the flow rate increases due to the decrease in system pressure loss. It is now supplied to the furnace 1.

Therefore, the reactor pressure decreases further, e.g.
/(i?, 11.), the second control device 44 is activated by a signal indicating that the reactor water level is low, and the valve 15.21 is activated.
is closed and the valve 38 of the second bypass line 36 is closed.
will be released. Therefore, the condensate discharged from the high-pressure condensate pump 13 passes through the check valves 37, 38, and 39'ft, and further through the first bypass pipe 35.
It is supplied to the nuclear reactor 1 through. In this case as well, since the supplied water bypasses the low-pressure water supply device 16, the water concentration further decreases and the flow rate also increases as described above.

When the reactor pressure further decreases to, for example, TK9/cTIt or lower, the second control device i4 further closes the valves 12.14 before and after the high-pressure condensate pump 13.
,.

9P40 of the eighth bypass pipe 42 is opened, and the condensate flowing out from the condensate desalination device 10 is transferred to the eighth bypass pipe 42.
, the second bypass line 36 , and the first bypass line 35 .

However, in this case as well, the water supply flow rate increases further due to the decrease in system pressure loss.

Changes in the water supply flow rate and the water supply temperature when the coolant is lost are shown in FIGS. 8 and 4 in the case of the present invention and in the case of the conventional case. That is, as shown in Fig. 8, when a coolant loss accident occurs and the water level drops to a certain extent, the main steam isolation valve 8 closes, the turbine-driven water supply pump loses its function, and the water supply flow rate decreases rapidly. Thereafter, it increases with the start of the motor-driven water pump. In the conventional device, as shown by the dotted line, a water supply flow rate of 50% of the constant heat # and amount is ensured. On the other hand, according to the present invention, since the first and second bypass pipes are opened in sequence, the water supply flow rate increases as described above. Moreover, as shown in FIG. 4, in the present invention, the supply water temperature also rapidly decreases after the accident occurs. As described above, according to the present invention, a large amount of low-temperature water can be supplied. Therefore, as shown in FIG. 5, the reactor pressure decreases more rapidly than in the past, and as a result, the flow rate of water injected into the emergency core cooling system also increases, as shown in FIG. 6.

As a result, as shown in Figure 7, the water level in the reactor shroud will recover faster and cool the fuel in the reactor core, and as shown in Figure 8, the peak of fuel cladding will be significantly lowered. This allows the reactor to maintain an even larger safety margin.

In the above embodiment, two control devices, the first and second, control the opening and closing of each valve, but as shown in FIG. It may be configured to function as a device. Further, the set value of the water level can be raised or lowered within a range that does not interfere with normal operation.

〔Effect of the invention〕

As explained above, according to the present invention, by adding a few valves, piping, and control devices to an existing system,
It is possible to inject a large amount of low-temperature feed water, speed up the depressurization of the reactor, and increase the total flow rate of river water in the emergency core cooling system. In addition, the water level in the g subreactor shroud is accelerated, and the fuel is rapidly cooled.
This has the effect of reducing the peak 1lIt of fuel cladding temperature and significantly improving the cooling performance of core fuel in the event of a loss of coolant accident.

[Brief explanation of drawings]

Fig. 1 is a system diagram of a conventional reactor water supply system, Fig. 2 is a system diagram of a water supply system according to an embodiment of the present invention, and Figs. Comparison chart of temperature, reactor pressure, cooling water flow rate of the emergency core cooling system, reactor water level, and cladding temperature over time between the conventional device and the device of the present invention, FIG. 9 is a diagram showing another implementation of the present invention. It is a system diagram showing an example. DESCRIPTION OF SYMBOLS 1... Nuclear reactor, 2... Main steam pipe, 13... High pressure condensate pump, 16... Low pressure feed water heater, 19... Turbine driven feed water pump, 23... High pressure feed water heater , 27
...′! lf@machine-driven water supply pump, 35... first bypass pipe line, 36... second bypass pipe line, 40.
...Eighth bypass pipeline, 43, required...control device. Applicant's agent Kiyotsuru Inomata Figure 2 - 477 - Figure 3 Figure 4

Claims (1)

[Claims] 1. A bypass pipe @1 that bypasses a high-pressure feedwater heater provided in a water supply system to a nuclear reactor, and a second bypass pipe that bypasses a low-pressure feedwater heater and a feedwater bonder. ,
an eighth bypass line that bypasses the high pressure condensate bonder;
The present invention is characterized in that it is provided with a control device that is activated by water level and pressure signals in the nuclear reactor and opens each of the bypass pipes. Boiling water nuclear power station water supply system. 2. The water supply system for a boiling water nuclear power plant according to claim 4, wherein the first bypass line is first opened when the water level in the reactor drops to a predetermined value. 3. The second bypass line and the eighth bypass line are:
The water supply device for a nail-shaped nuclear power plant according to claim 1 or 2 is configured to open sequentially in response to a low water level and a low pressure signal in the reactor. .