JPS62144093A - Protective device for nuclear 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 [Industrial Application Field] The present invention relates to a nuclear reactor protection device, and particularly to a nuclear reactor protection device that is suitable for application to a nuclear reactor having an internal pump.

[Prior Art] A new type of boiling water reactor has been developed to replace the boiling water reactor in which the core flow rate is adjusted 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. 59-84197 and Japanese Patent Application Laid-open No. 59-18859 are known.
9 and Japanese Unexamined Patent Publication No. 60-15599 are known.

FIG. 4 shows a boiling water reactor with a conventional internal pump. The internal pump 21 is installed within a reactor pressure vessel 23 containing a reactor core 22. A rotating shaft of the internal pump 21 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.

[Problems to be solved by the invention] In a conventional nuclear reactor that includes multiple internal pumps, assuming that all the internal pumps are tripped, there is a possibility that the cooling capacity of the core will decrease transiently. I newly discovered that there is.

That is, the combined inertia of the internal pump 21 and the static inverter 19 in a nuclear reactor with an internal pump is the combined inertia of the recirculation pump and the MG upper 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
When all the internal pumps 21 trip due to a loss of power to the reactor 8, the rotational speed of the internal pumps rapidly decreases by 1; and the core flow rate rapidly decreases. 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 nuclear reactor protection device that can scram a nuclear reactor in a short time even when all internal pumps trip.

[Means for Solving the Problems] The above problem is such that the flow rate of coolant supplied to the core as measured by a flow meter is equal to or higher than a first predetermined level, and within a predetermined time, the flow rate of coolant is increased to a second level. This can be solved by providing a scram determination means that outputs a scram signal when the scram determination means is below a predetermined level, and rapidly inserting the control rods into the reactor core based on the scram signal output from the scram determination means.

[Operation] The coscrum judgment device judges whether or not a reactor scram is necessary due to a sudden reduction in the core flow rate due to all trips of the internal pumps, and if necessary, it is output as a scram signal, and based on this scram signal, the reactor is activated. Scrammed.

[Example 1] A preferred embodiment of the present invention applied to a boiling water reactor is shown below.
This will be explained based on FIGS. 1 and 2.

Boiling water reactors with internal pumps have core 2
An internal pump 21 is installed in a reactor pressure vessel 23 containing a reactor pressure vessel 23. 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 installed outside the reactor pressure vessel 23 and 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 for one internal pump 21
are concatenated. 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. control rod 16
is connected to the control rod drive device 15. A flow meter 28 is provided to determine the core flow rate. Reference numeral 19 denotes a static inverter, which is connected to the mother aL8, which is a power source. 20 is a recirculation flow rate control device.

The reactor protection device of this embodiment includes a control rod drive control device 3, a scram determination device 17, and a flowmeter 28. Details of the scrum judgment device 17 will be explained based on FIG. 2. The scram determination device 17 includes the initial core flow rate determination section 4
, a core flow rate determining section 11 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 . AND circuit 29
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 composed of a scram valve, a scram outlet valve, and a scram pilot solenoid valve as disclosed in Japanese Patent Application Laid-Open No. 51-137091. AND circuit 2
The output signal 9 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 results will be explained below based on the characteristics shown in FIG.

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 of the stationary inverter, and the reactor core flow rate suddenly decreases. This is an event that could result in a rapid and transient decrease in core cooling capacity. Therefore, when considering such a protection 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 adverse effect on the reactor due to a sudden decrease in core flow rate.

On the other hand, if the initial core flow rate is low, the reduction in core cooling capacity will not be a problem even if the internal pump trips.

According to the analysis, the lower limit of initial core flow rate at which adverse effects occur is 5.
It is considered to be 5% flow rate.

(2) The larger and more rapid the fall in the core flow rate, the greater the negative impact, but fluctuations in the 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 by more than 50%/sec.

(3) Care should be taken to avoid issuing unnecessary scram signals during normal operation or when starting and stopping.

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

(a) When the initial core flow rate is approximately 55% or more, 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 a natural circulation state is reached when all recirculation pumps are tripped. 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 the "abnormal capacity".

(C) Perform reactor protection operations when the rate of decrease in core flow rate is faster than approximately 50%/sec.

The nuclear 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 sent to the filter 10.
After the noise is removed, the signal is 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 compares the core flow rate signal output from the filter 1o with the initial core flow rate determination value (55% flow rate) 5, and when the former level exceeds the latter level, the initial core flow rate determination unit 4 determines the initial core flow rate determination value. A signal 6 indicating that the value is greater than or equal to the value is output. This signal 6 is stored in the signal holding circuit 7 for a predetermined time (
After being held (for example, for about 2 seconds), the signal holding circuit 7 outputs the signal 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 core flow rate decrease rate is 50%. /second, and does not pose any 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 the core flow rate determination unit (approximately 35% flow rate) 12, and if the former level is lower than the latter level, the core flow rate determination unit 11 determines that the core flow rate is low. Outputs signal 13. This “core flow rate low” signal 13 is input to the AND circuit 29.

The AND circuit 29 determines that the initial core flow rate is a predetermined value (judgment value: 5
The holding circuit 7 holds the "initial core flow rate judgment value or higher" signal 6 for 2 seconds, indicating that the current core flow rate is below a predetermined value (judgment value of 35%). When the "core flow rate low" signal 13 is input, the core flow rate sudden decrease scram signal 14 is output.

That is, it is determined that the core flow rate has decreased to 35% or less within 2 seconds from the plant state where the core flow rate is 55% or more, and the scram signal 14 is 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 operates as shown in Japanese Patent Application Laid-Open No. 137091/1980 when the scram signal 14 is input, and opens the scram population 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 we hypothetically assume that all the internal pumps 21 trip, 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 a turbine trip. If a turbine trip occurs, the reactor will scram and shut down safely. However, this embodiment can scram the reactor in a shorter time than such a reactor protection device.

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.

[Effects 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.

[Brief explanation of drawings]

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 of the Kerr core flow rate curve out of the reactor explaining the logic of this, and is a configuration diagram of a nuclear reactor having an internal pump. 4... Initial core flow rate determination section, 7... Signal holding section, 1
゜... Filter, 11... Core flow rate determination section, 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 a control device inserted into the reactor core. A nuclear reactor protection device having a control rod and means for driving the control rod, wherein the coolant flow rate measured by a flow meter that detects the flow rate of the coolant supplied to the reactor core is equal to or higher than a first predetermined level. scram determining means for outputting a scram signal when the coolant flow rate falls below a second predetermined level within a predetermined time; and means for controlling the control rod drive device to cause rapid insertion. 2. Claim 1, wherein the first predetermined level is 55% flow rate and the second predetermined level is 35% flow rate.
Reactor protection device as described in Section 1.