JPS648316B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to an emergency core cooling system for a boiling water nuclear power plant. The emergency core cooling system flow rate is controlled to follow changes in the coolant outflow volume (relief safety valve flow rate) to control the reactor water level to an appropriate position, and reduce the probability of failure due to repeated startup and shutdown of the core cooling system. related to the control of the emergency core cooling system.

In the case of a conventional emergency core cooling system control system, in the event of a loss of coolant accident, the reactor water level drops rapidly, and a signal indicating that the reactor water level is low activates the emergency core cooling system and injects cooling water into the reactor core.

In this case, the core cooling system control device controls the discharge flow rate of the variable speed pump to the rated flow rate.

In particular, when the relief safety valve fails to open, the coolant in the reactor pressure vessel flows into the pressure control chamber through the relief relief valve, and as shown by the solid line J in Figure 4, the water level drops rapidly from the normal operating water level W. When point a is reached, the emergency core cooling system is activated and cooling water is injected into the reactor pressure vessel as shown by solid line U in FIG. However, in this case, the injection amount is controlled (constant value control) to maintain the pump discharge flow rate as described above.
The difference between the cooling water injection amount and the relief valve flow rate contributes to the water level rise. In other words, when the emergency core cooling system is activated,
The reactor water level rises from point a in Figure 4, exceeds the normal operating water level at point b, and rises further.This is installed to prevent overflow of supercharged water into the main steam line and adverse effects on the turbine side due to increased carryover. Trip point c is reached due to the high water level. When the reactor water level reaches this trip point c, the emergency core cooling system is shut down and the flow rate decreases from point g in FIG. 5.
Since cooling water is not injected into the reactor pressure vessel by the trip, the reactor water level returns to the trip point c.
It further decreases and reaches point d, which is at the same level as point a. At this point d, the emergency core cooling system is restarted again. (Point h in Figure 5) From then on, the above steps of starting and stopping will be repeated. For this reason, the number of steps for starting and stopping increases, and the probability of restart failure becomes extremely high.

On the other hand, the reactor pressure decrease due to the sudden opening of the relief safety valve causes the reactor water to flash, and the water level gauge connected to the reactor pressure vessel may no longer indicate an appropriate water level. Furthermore, due to the apparent rise in water level due to this erroneous water level measurement, there is a possibility that the emergency core cooling system will be shut down due to operator error at the above-mentioned high water level trip point.

The present invention was made in view of the above circumstances, and
The purpose of this is to maintain the reactor water level appropriately in the event of a loss of coolant accident due to an open failure of a relief valve, to prevent malfunctions due to repeated startup and shutdown of the emergency core cooling system, and to monitor the reactor water level gauge. An object of the present invention is to obtain an emergency core cooling system control device capable of preventing erroneous operations by operators due to malfunctions.

An embodiment of the present invention will be described below with reference to the drawings. In FIG. 1, steam generated within a reactor pressure vessel 1 is sent to a turbine (not shown) through a main steam pipe 3, and is converted into electrical energy by a generator.

The main steam pipe 3 is equipped with over ten relief safety valves 4 (abbreviated to one in the figure) in order to maintain the integrity of the pressure boundary when the pressure inside the reactor pressure vessel 1 increases.
It is provided. The steam released from the relief safety valve 4 passes through the exhaust pipe 5, is led to the pressure suppression chamber 6, and is condensed. Furthermore, a main steam isolation valve 7 is provided in the main steam pipe 3 to maintain the integrity of the reactor primary system.
This main steam isolation valve 7 is fully closed at the start-up water level of the emergency core cooling system 30.

On the other hand, the emergency core cooling system 30 is composed of a water source and a driving pump. The water source is condensate storage tank 8
Alternatively, using the pressure suppression chamber 6, the water level in each water chamber automatically switches to the water supply line 10a connected to the condensate storage tank 8 shown by a solid line or the water supply line 10b connected to the pressure suppression chamber 6 shown by a broken line. Ru.

Water from the water source is pressurized by the emergency core cooling system pump 9 and is injected into the reactor pressure vessel 1 via the water supply line 10 . The control device for the emergency core cooling system 30 uses an exhaust pipe (not shown in the figure) connected to another relief safety valve 4 that is similar to the flow rate measured by a flow meter 13 attached to an exhaust pipe 5 connected to the relief safety valve 4. ), the safety relief valve total flow rate signal W _s , 14a is obtained by adding the flow rate measured by the flow meter attached to the pump 14 by the adder 14, and the flow meter 15 disposed on the discharge side of the emergency core cooling system pump 9. Comparison calculation circuit 1 which inputs the emergency core cooling system flow rate signal W _E , 15a
2, the emergency core cooling system pump control device 11
The output signal 12a is transmitted to the emergency core cooling system 30 to control the flow rate of the emergency core cooling system 30.

Now, the details of the comparison calculation circuit 12 are as shown in FIG. First, the contact 22 shown in the relay sequence in Figure 3 is turned “ON” by the opening operation of the relief safety valve.
Then, the relay K23 is energized, and the integrator 16 shown in FIG. 2 starts integrating the safety relief valve total flow signal W _S , 14a. When the emergency core cooling system is activated, the contact 24 in Fig. 3 turns "ON" and the integrator 1
The integrator output signal 16a, which is obtained by integrating the time _t1 from the start of integration of step 6 until the relay L25 is energized, is sent to the adder 1.
8 is input. Simultaneously with the excitation of this relay 25, the integrator 17 generates a relief safety valve total flow signal W _S , 14
a starts to be integrated, and the integrator output signal 17a is input to the adder 18. On the other hand, the emergency core cooling system flow rate signal W _E , 15a is also integrated by the integrator 19 together with the excitation of the relay L25, and the integrator output signal 19a and the adder output signal 18a are sent to the comparator 2.
Compare by 0. The integrators 17 and 19 integrate until the value of the comparator 20, that is, the difference between the integrator output signal 19a and the adder output signal 18a becomes zero (time _t2 ). When the value of comparator 20 is greater than zero, contact 26 in FIG. 3 turns "ON" and relay M27 is energized. Therefore, the contact P28 in FIG. 2 is turned "ON" by the excitation of the relay 27, and during this time, the set value of the emergency core cooling system controller 21 becomes the adder output signal 18a output from the adder 18, and the emergency core cooling system is activated. The pump 9 of the emergency core cooling system 30 is controlled so that the flow rate signal 15a becomes the adder output signal 18a.

Next, when the value of comparator 20 becomes zero, contact 26 in FIG. 3 turns "OFF", relay M27 becomes de-energized, contact P28 in FIG. 2 turns "OFF", and relay N29
becomes “ON”. Then, the controller 21 is set to the total flow rate of the safety relief valves, and the output signal 12a is sent to the pump 9 via the emergency core cooling system controller 21 so that the flow rate signal 15a of the emergency core cooling system becomes the total flow rate signal 14a of the safety relief valves. This pump 9 is controlled.

When an abnormal transient phenomenon occurs in which the pressure inside the reactor pressure vessel 1 increases, the relief safety valve 4 is opened and the pressure inside the reactor is released to the pressure suppression chamber 6. At this time, the relief safety valve 4 fails to open (stack open).
In this case, the cooling water 2 in the reactor pressure vessel 1 is
As shown in the figure, the water level decreases from the normal operating water level W and reaches point a. The emergency core cooling system 30 injects cooling water into the reactor pressure vessel 1 from point f in FIG. Simultaneously with the opening operation of the relief safety valve 4, the contact 22 is turned ON, the relay K23 is energized, the emergency core cooling system 30 is activated, the contact 24 is turned ON, and the integrator 16 is kept under the reactor pressure until the relay L25 is energized. Integrate the flow rate from container 1 (total flow rate of safety relief valve). That is, the sixth
Corresponds to part Q ₂ in the diagram.

When the contact 24 turns ON and the relay L25 is energized, the integrator 17 outputs the relief safety valve total flow signal 14a.
The integrated value and the outflow amount Q ₂ are input to the adder 18. The adder output signal 18a output from the integrator 17 is compared with the integrator output signal 19a of the emergency core cooling system flow rate signal 15a via the integrator 19 by a comparator 20, and the above-mentioned integral is applied until the difference becomes zero. 17 and 19 perform integration.

In other words, the emergency core cooling system flow rate shown in Figure 5 is reduced by the amount corresponding to the sum of the total amount of safety relief valves Q ₂ and R ₂ in Figure 6.
We will inject Q ₁ and R ₁ .

During this period, the reactor water level W reaches the normal operating water level from point a to point b, as shown in FIG.

When the signal of comparator 20 becomes zero, relay P28 in FIG. 2 turns "OFF" and relay N29 turns "ON".
Therefore, the emergency core cooling system flow rate is controlled to be equal to the relief safety valve total flow rate signal 14a.

As shown in FIG. 5, the emergency core cooling system flow rate is as indicated by a broken line 99.

That is, the same cooling water S ₁ (shown in FIG. 5) as the safety relief valve flow rate S ₂ shown in FIG. 6 will be injected into the reactor pressure vessel 1, and the water level of the reactor water 2 will be As shown in Figure 4, from point b to dotted line i
The water level is maintained at the normal operating water level as shown in .

Thereafter, if the safety relief valve total flow rate decreases, the emergency core cooling system flow rate also decreases, making it possible to control the outflow rate from the reactor pressure vessel and the injection rate to become the same. Therefore, the reactor water level is kept constant, and even if a malfunction occurs in the water level instrumentation system and monitoring becomes impossible, the reactor water inventory is maintained and the integrity of the fuel cladding is maintained sufficiently. It turns out.

Thus, in a boiling water nuclear power plant incorporating the emergency core cooling system control device according to the present invention, the reactor water level can be controlled to an appropriate position in the event of a safety relief valve opening failure, and the core cooling system can be controlled to an appropriate position. It is possible to reduce the probability of failure due to repeated starting and stopping of nuclear reactors, and maintain the integrity of the fuel cladding inside the reactor.

[Brief explanation of drawings]

Figure 1 is a system diagram showing one embodiment of the present invention, Figure 2 is a system diagram showing an embodiment of the present invention.
, 3 are block diagrams showing a comparison calculation circuit in one embodiment of the present invention, and FIGS. 4, 5, and 6 show the reactor water level, emergency core cooling water flow rate, and and a characteristic diagram for explaining the behavior of the relief safety valve flow rate. 1... Reactor pressure vessel, 2... Safety relief valve, 12...
Comparison calculation circuit, 30...Emergency core cooling system.

Claims (1)

[Claims] 1 In the emergency core cooling system control device that supplies cooling water to the reactor pressure vessel in the event of a loss of coolant accident in the reactor pressure vessel of a boiling water nuclear power plant, the main steam pipe connected to the reactor pressure vessel If a loss of coolant accident occurs due to an open failure of the relief safety valve installed in the reactor pressure vessel, the amount of coolant that leaked out from the reactor pressure vessel through the relief safety valve and the cooling water injected from the emergency core cooling system. 1. A control device for an emergency core cooling system, characterized in that the amount of injection is compared and calculated, and the amount of injection into the emergency core cooling system is controlled so that the value obtained by the comparison calculation becomes zero.