JPH0518396B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION [Field of Industrial Application] The present invention relates to a nuclear reactor protection device, and particularly to a nuclear reactor protection device suitable for application to a nuclear reactor having an internal pump.

[Conventional technology]

A new type of boiling water reactor is being developed to replace the boiling water reactor, in which the core flow rate is controlled by a recirculation pump installed in the recirculation system piping. In this boiling water reactor, the core flow rate is adjusted by an internal pump provided within the reactor pressure vessel. As a protection device for a boiling water reactor having such an internal pump, Japanese Patent Application Laid-open No. 59
Those shown in Japanese Patent Application Laid-open No. 84197, Japanese Patent Application Laid-Open No. 59-188599, and Japanese Patent Application Laid-open No. 60-15599 are known.

FIG. 4 shows a boiling water reactor with a conventional internal pump. The internal pump 21 is a reactor pressure vessel 2 that contains a reactor core 22.
It is located within 3. Internal pump 21
The rotating shaft of is connected to a motor 24 installed outside the reactor pressure vessel 23. The motor 24 of the internal pump 21 is supplied with power from the static inverter 19 . The restraint type inverter 19 is
The power frequency input from the bus bar 18 is adjusted based on the control signal output from the recirculation flow rate control device 20, and the rotational speed of the internal pump 21 is controlled.

[Problem to be solved by the invention] In a conventional nuclear reactor with multiple internal pumps, if all internal pumps are hypothetically tripped, there is a possibility that the cooling capacity of the core will decrease transiently. I newly found out that there is.

That is, the combined inertia of the internal pump 21 and stationary inverter 19 in a nuclear reactor with an internal pump is the combined inertia of the recirculation pump and the MG set that controls its rotation speed in a nuclear reactor with recirculation system piping. becomes very small compared to . For this reason, nuclear reactors with internal pumps are superior in quick response to requests for changes in reactor output, etc. However, in the unlikely event that bus 1
If all the internal pumps 21 trip due to a loss of power to the reactor 8, the rotational speed of the internal pumps will drop rapidly and the core flow rate will drop sharply. Such a sudden decrease in the core flow rate leads to a sudden decrease in the core cooling capacity, which may lead to an unfavorable state from the viewpoint of the thermal margin of the fuel.

An object of the present invention is to provide a reactor protection device that can avoid unnecessary scrams even when all internal pumps trip, and that can scram the reactor in a short time when all internal pumps trip and a scram is necessary. Our goal is to provide the following.

[Means for solving problems]

It is an object of the present invention to provide a system in which the coolant flow rate measured by a flow meter that detects the flow rate of coolant supplied to the core is equal to or higher than a first predetermined flow rate when all internal pumps are tripped; outputting a scram signal when the coolant flow rate becomes equal to or less than a second predetermined flow rate, which is smaller than the first predetermined flow rate, after a predetermined time has elapsed from the time when all the null pumps trip;
scram determination means for not outputting the scram signal when the coolant flow rate when the internal pumps are all tripped is smaller than the first predetermined flow rate; This can be achieved by comprising means for controlling the control rod drive device to achieve rapid insertion.

[Function] The coolant flow rate is equal to or higher than the first predetermined flow rate when all the internal pumps are tripped, and the coolant flow rate is smaller than the first predetermined flow rate when a predetermined time has elapsed from the time when all the internal pumps are tripped. Since the reactor is scrammed by the scram signal that is output when the flow rate falls below the second predetermined flow rate, the reactor can be scrammed in a short time.

Further, if the coolant flow rate when all internal pumps are tripped is smaller than the first predetermined flow rate, a scram of the nuclear reactor can be avoided.
This reactor scram should also be avoided if, even if the coolant flow rate when all internal pumps are tripped is above the first predetermined flow rate, the coolant flow rate does not become below the second predetermined flow rate after the elapse of the above predetermined time. I can do it. In this way, even when all internal pumps trip, unnecessary reactor scrams can be avoided.

〔Example〕

A preferred embodiment of the present invention applied to a boiling water reactor will be described with reference to FIGS. 1 and 2.
A boiling water reactor with an internal pump is
An internal pump 21 is installed within a reactor pressure vessel 23 containing a reactor core 22. The ten internal pumps 21 are arranged in an annular gap between the reactor pressure vessel 23 and the core shroud 25 surrounding the reactor core 22 so as to surround the reactor core shroud 25 . Ten motors 24 are connected to the reactor pressure vessel 2.
3 and installed at the bottom of the reactor pressure vessel 23. A rotating shaft of the internal pump 21 passes through the lower wall of the reactor pressure vessel 23 and is connected to the motor 24 . One motor 24 is connected to one internal pump 21 . A control rod 16 that adjusts the reactor output is installed so that it can be moved in and out of the reactor core 22. The control rod 16 is connected to the control rod drive device 1
It is connected to 5. A flow meter 28 is provided to measure the core flow rate. 19 is a static inverter connected to a bus 18 which is a power source.
20 is a recirculation flow control device.

The reactor protection device of this embodiment includes a control rod drive control device 3, a scram determination device 17, and a flow meter 2.
It has 8. Details of the scrum judgment device 17 will be explained based on FIG. 2. Scrum judgment device 1
7 is an initial core flow rate determination unit 4, a core flow rate determination unit 1;
1 and an AND circuit 29. 7 is a signal holding section, which is a kind of delay circuit. The initial core flow rate determining section 4 and the core flow rate determining section 11 are connected to a flow meter 28 via a filter 10. One input end of the AND circuit 29 is connected to the initial core flow rate determining section 4 via the signal holding section 7, and the other input end of the AND circuit 29 is connected to the core flow rate determining section 11. The output end of the AND circuit 29 is connected to the control rod drive device control device 3. The control rod drive control device 3 is comprised of a scram inlet valve, a scram outlet valve, and a scram pilot solenoid valve as disclosed in Japanese Patent Laid-Open No. 51-137091. The output signal of the AND circuit 29 becomes an opening signal for the scram pilot solenoid valve.

The reactor protection device of this example was installed in a nuclear reactor based on the following study results. The result is the third
This will be explained below based on the characteristics of the figure.

The event targeted by the reactor protection system of this embodiment is a hypothetical event in which all internal pumps trip due to a loss of power to the stationary inverter 19 or a trip in the stationary inverter, and a sudden decrease in the core flow rate. This is an event in which the core cooling capacity may suddenly decrease temporarily.
Therefore, when considering the protective device, it is necessary to consider the following three points.

(1) The higher the core flow rate at the time when all internal pumps trip (hereinafter referred to as initial core flow rate), the greater the negative impact on the reactor due to a sudden decrease in core flow rate. On the other hand, if the initial core flow rate is low, even if the internal pump trips, the reduction in core cooling capacity will not be a problem. According to the analysis, the lower limit of initial core flow rate at which adverse effects occur is 55
% flow rate.

(2) The larger and more rapid the decrease in core flow rate, the greater the negative impact, but fluctuations in core flow rate within the normal operating range above the internal pump minimum speed operating point are not a problem, and all internal pumps trip. It is sufficient to protect against a hypothetical event in which the core flow rate suddenly and significantly decreases for more than 50% seconds.

(3) Care should be taken to avoid issuing unnecessary scram signals during normal operation or during startup and stop.

The logic of reactor protection considering the above points is as follows.

(a) When the initial core flow rate is approximately 55% or higher, perform reactor protection operations, that is, scram the reactor.

(b) The range up to the lowest pump speed line is the range in which operation is expected during normal operation, and when all recirculation pumps trip, a natural circulation state is reached. Therefore, it is reasonable to scram between the natural circulation line and the lowest pump speed line.
That is, the reactor protection operation is performed when the core flow rate is below about 35%, which is "abnormally low."

(c) If the rate of decrease in core flow rate is faster than approximately 50%/sec, perform reactor protection operations.

The reactor protection device shown in FIG. 1 has the functions (a) to (c) above.

The core flow rate signal 9 measured by the flow meter 28 is input to the filter 10 to remove noise, and then input to the initial core flow rate determining section 4 and the core flow rate determining section 11.

The initial core flow rate determination unit 4 and the core flow rate signal output from the filter 10 and the initial core flow rate determination value (55%
When the former level exceeds the latter level, a signal 6 indicating that the flow rate is "initial core flow rate determination value or higher" is output. This signal 6
is held in the signal holding circuit 7 for a predetermined period of time (for example, about 2 seconds) and then output from the signal holding circuit 7 to the AND circuit 29. If the signal output from the core flow rate determination unit 11 becomes low core flow rate after 2 seconds or more have passed since the initial core flow rate signal exceeded the initial core flow rate determination value 5, the rate of decrease in the core flow rate will be
This is sufficiently slow compared to 50%/sec, and does not pose a problem in terms of reactor protection measures.

On the other hand, the core flow rate determination unit 11 compares the core flow rate signal with a core flow rate low determination value (approximately 35% flow rate) 12, and if the former level becomes smaller than the latter level, the core flow rate determination unit 11 determines that the core flow rate is low. The signal 13 is output. This “core flow rate low” signal 13 is input to the AND circuit 29.

The AND circuit 29 outputs the signal 6, which indicates that the initial core flow rate is equal to or higher than a predetermined value (judgment value: 55%), held for 2 seconds by the holding circuit 7, and the current core flow rate. is a predetermined value (judgment value 35
When the “core flow rate low” signal 13 indicating that the core flow rate is below %) is input, the core flow rate suddenly decreases scram signal 1
Outputs 4.

That is, it is determined that the core flow rate has decreased to 35% or less after 2 seconds have elapsed from the plant state where the core flow rate is 55% or more, and the scram signal 14 is output. This scram judgment device 17 determines whether the initial core flow rate is smaller than a predetermined value (55% flow rate) or even when the initial core flow rate is above the predetermined value (55% flow rate), from the time when all internal pumps trip. When the core flow rate does not fall below the low judgment value (35% flow rate) after a predetermined time (2 seconds) has elapsed, the scram signal 14 is not output.

The core flow rate sudden decrease scram signal 14 is input to the scram pilot solenoid valve of the control rod drive device control device 3. The scram pilot electromagnetic valve is actuated by input of the scram signal 14 as shown in Japanese Patent Application Laid-Open No. 137091/1982 to open the scram inlet valve and the scram outlet valve. Therefore, high-pressure driving water is supplied from the accumulator to the control rod drive device 15, and the control rod 16 is rapidly inserted into the reactor core 22 by driving the control rod drive device 15. Therefore, the reactor is scrammed.

In the system currently being considered, if all the internal pumps 21 are hypothetically tripped, the amount of voids in the core 22 will increase rapidly due to a sudden decrease in the core flow rate, and the reactor water level will rise. leading to turbine trip. When a turbine trip occurs, the reactor scrams and shuts down safely. However, this embodiment can scram the reactor in a shorter time than such a reactor protection device.

As described above, in this embodiment, when the initial core flow rate is smaller than a predetermined value, and even if the initial core flow rate is greater than or equal to the predetermined value, the core flow rate is determined at the predetermined time point from the time when all internal pumps trip. Since the scram signal 14 is not output when the voltage does not fall below the low judgment value, unnecessary scrams can be avoided even when all internal pumps trip.

It should be noted that the judgment values and signal holding times shown in this embodiment are representative examples, and are set appropriately for each plant.

〔Effect of the invention〕

According to the present invention, even if all internal pumps trip, the reactor can be scrammed in a short time, and the safety of the reactor is significantly improved.

Furthermore, even if the coolant flow rate when all the internal pumps are tripped is smaller than the first predetermined flow rate, and even if the coolant flow rate when all the internal pumps are tripped is greater than or equal to the first predetermined flow rate, the predetermined time period elapses. Unnecessary scrams can be avoided even when all internal pumps trip when the coolant flow rate does not fall below the second predetermined flow rate.

[Brief explanation of the drawing]

Fig. 1 is a block diagram of a nuclear reactor protection device that is a preferred embodiment of the present invention, Fig. 2 is a detailed block diagram of the scram judgment device of Fig. 1, and Fig. 3 is a block diagram of the reactor protection device of Fig. 1. FIG. 4 is a diagram showing a reactor power-core flow rate curve explaining the logic of this, and is a configuration diagram of a nuclear reactor having an internal pump. 4... Initial core flow rate determination unit, 7... Signal holding unit, 10... Filter, 11... Core flow rate determination unit, 15... Control rod drive device, 16... Control rod,
17...Scram judgment device, 21...Internal pump, 22...Reactor core, 24...Motor, 29
...AND circuit.

Claims (1)

[Scope of Claims] 1. A reactor vessel containing a reactor core, a plurality of internal pumps provided in the reactor vessel and supplying coolant into the reactor core, and control rods inserted into the reactor core. and a device for driving the control rods, wherein the coolant flow rate measured by a flow meter that detects the flow rate of the coolant supplied to the reactor core is determined when the internal pumps are fully tripped. When the coolant flow rate is equal to or higher than the first predetermined flow rate and becomes equal to or lower than the second predetermined flow rate, which is smaller than the first predetermined flow rate, after a predetermined time has elapsed from the time when all the internal pumps are tripped, the scram occurs. scram determination means that outputs a signal and does not output the scram signal when the coolant flow rate when the internal pump is fully tripped is smaller than the first predetermined flow rate; A nuclear reactor protection device comprising means for controlling the control rod drive device to rapidly insert control rods into the reactor core. 2. Claim 1, wherein the first predetermined flow rate is a 55% flow rate, and the second predetermined flow rate is a 35% flow rate.
Reactor protection device as described in Section 1.