JPS62115396A - 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 is directed to a new type of boiling water reactor (hereinafter referred to as A-BWR).
The present invention relates to a nuclear reactor protection device that protects a nuclear reactor even when, for example, a coolant circulation pump that forcibly circulates coolant into the reactor core stops and the reactor core flow rate suddenly decreases.

[Prior Art] A conventional example will be explained with reference to FIG. Figure 3 is A-BW
1 is a cross-sectional view showing a schematic configuration of R, and reference numeral 1 in the figure is a reactor pressure vessel. Inside this reactor pressure vessel 1 is a coolant 2.
and a reactor core 3 are housed therein. This core 3 is composed of a plurality of fuel assemblies and control rods 4 (only one is shown in the figure), which are not shown. The coolant 2 flows upward through the core 3 and is heated by the heat of nuclear reaction at 3° in the core. The heated coolant 2 enters a two-phase flow state of water and steam. The coolant 2 in this two-phase flow state is introduced into a steam/water separator 5 installed above the reactor core 3 and separated into steam and water. The separated internal steam is introduced into a steam dryer 6 installed above the steam separator 5 and dried to become dry steam. This dry steam is placed in the main steam connected to the reactor pressure vessel 1! It is transferred to a turbine system (not shown) via 12 and used for power generation. On the other hand, the separated internal water is in the reactor pressure vessel 1.
The water flows down the downcomer section 8 between the shroud 7 and the water supply pipe 13 and joins with the water supply flowing in through the water supply sparger 9, and is sucked into a recirculation pump (hereinafter referred to as an internal pump) 11. This internal pump 11
The fuel is then pressurized and supplied to the lower part of the reactor core 3 again. Incidentally, a plurality of the internal pumps 11 are installed in the circumferential direction,
Each is driven by a motor.

A safety protection device 21 is installed in the A-BWR having such a configuration. For example, the pressure increase in the reactor pressure vessel 1,
Alternatively, when an abnormality such as a drop in the water level occurs, the occurrence of a v4 abnormal situation is detected and an abnormality occurrence signal S22 is output to the safety protection device 21. As a result, the safety protection device 21 outputs a control signal 824 to the control rod drive mechanism 23, causing the control rod 4 to be urgently inserted into the reactor core 3, and rapidly reducing the reactor power. This is the so-called scrum operation.

[Problems with the paradoxical technology] In the above configuration, the internal pump 11 is not suitable for the case where two recirculation pumps are installed outside the reactor pressure vessel 1 (BW
It is smaller than the recirculation pump of the R type), and in the event of a loss of pump power, the flow rate is expected to drop rapidly due to the small inertia of the small pump, resulting in a rapid decrease in the core flow rate. Ru. In such a case, there is a possibility that the heat generated within the reactor core 3 will not be removed sufficiently. This also applies when the internal pump 11 is further downsized as the plant becomes smaller. Therefore, in order to prevent such a situation from occurring in the past, the motor power supply for the internal pump 11 is made sufficiently reliable, and the power supply system is divided into degrees, and the motor power supply for all internal pumps 11 is This ensures that power is not lost at the same time.

However, in order to further improve safety, it is necessary to maintain the integrity of the core 3 even in the event that the motor power of all internal pumps 11 is lost. It was requested.

[Objective of the Invention] The present invention has been made based on the above points, and its purpose is to prevent the core flow rate from suddenly decreasing in the event that the motor power of all internal pumps is lost. An object of the present invention is to provide a nuclear reactor protection device that can maintain the integrity of the fuel and the integrity of the reactor core even if such a situation occurs.

[Summary of the Invention] That is, the nuclear reactor protection device according to the present invention is characterized in that it monitors the core flow rate of a nuclear reactor, detects that the core flow rate has decreased abnormally, and causes an emergency shutdown of the reactor. be.

In other words, the reactor core flow rate is constantly monitored and if the core flow rate l drops abnormally, the reactor is brought to an emergency shutdown.

[Embodiment of the Invention] An embodiment of the present invention will be described below with reference to FIGS. 1 and 2. Note that the same parts as in the prior art will be described with the same reference numerals. Figure 1 shows the reactor protection system W1 according to this embodiment.
21, in which reference numeral 131 is a filter circuit. A core flow rate signal 5132 is input to this filter circuit 131 . The filter circuit 131 removes minute fluctuations and signal noise during normal operation from the core flow rate signal 5132. Generally, the period of minute fluctuations and signal noise during normal operation is relatively short, for example 0.1 to 0.
．． The time constant is about 5 seconds, and therefore, the filter circuit 131 having a time constant of about 5 seconds is also used in this embodiment. Then, the flow rate signal 5132 from which minute fluctuations and signal noise during normal operation have been removed by the filter circuit 131 is sent to the differential circuit 131.
3 is input. This differentiation circuit 133 calculates the rate of decrease in flow rate.

The core flow rate reduction rate signal 5134 is then input to the comparison circuit 135. On the other hand, a preset core flow rate reduction rate setting signal 8136 is input to this comparison circuit 135. At that time, the rate of decrease in the core flow rate, at which the integrity of the fuel is a problem, is (
30%/sec) or more, and therefore, this value is used as the core flow rate setting signal 8136. The comparison circuit 135 outputs a scram signal 8138 to the scram system 137 when the core flow rate reduction rate signal 5134 exceeds the core flow rate reduction rate setting signal 8136. As a result, the scram system 137 is activated and the control rods 4 are moved to the reactor core 3.
It was inserted into the reactor, and the power of the reactor core was rapidly reduced.

As described above, according to this embodiment, even if, for example, a situation occurs in which the motor power of all internal pumps 11 is lost and the core flow rate suddenly decreases, this is detected and the reactor is scrammed to reduce the core output. Since it can be rapidly reduced, it is possible to reliably maintain the integrity of the entire nuclear reactor as well as the integrity of the fuel and core, making it possible to significantly improve safety. This will be shown in comparison with the conventional method.

In Figure 2, the horizontal axis shows time, and the vertical axis shows the fuel rod cladding 'a.
This is a diagram showing the change in the cladding tube temperature over time, in which the broken line shows the case of the conventional case and the solid line shows the case of the present example. As is clear from FIG. 2, it can be seen that in the case of this example, the temperature rise in the cladding tube of the fuel rod is effectively suppressed.

Note that the present invention is not limited to the above-mentioned embodiments, and various embodiments are possible. For example, in the embodiment described above, a differential circuit is used to calculate the rate of decrease in core flow rate, but the present invention is not limited to this, and any other circuit may be used as long as it can detect the rate of decrease in core am. Furthermore, in the above embodiment, the problem was the rate of decrease in the core flow rate, but if the core flow rate signal and the core output signal are extracted and the difference between the two is calculated, and the difference exceeds a set value (the core flow rate is A configuration may also be used in which the scram signal is output when the signal is too small. In addition to this, a configuration may be adopted in which a core flow rate signal is extracted, a percentage of the rated value is calculated, and a scram signal is output when the value is less than a set value.

[Effects of the Invention] As detailed above, according to the nuclear reactor protection device according to the present invention, even if, for example, the motor power of all internal pumps is lost and the reactor core flow rate suddenly decreases, this can be reliably detected. Since the nuclear reactor can be brought to an emergency shutdown, the integrity of the fuel rods and reactor core, and by extension the integrity of the reactor, can be maintained reliably.

[Brief explanation of drawings]

Figures 1 and 2 are diagrams showing one embodiment of the present invention. Figure 1 is a diagram showing the configuration of a nuclear reactor protection device, and Figure 2 is a diagram showing the temporal change in temperature of the fuel rod cladding compared to the conventional one. Comparison diagram, 3rd
The figure shows the configuration of a conventional boiling water reactor. DESCRIPTION OF SYMBOLS 1... Reactor pressure vessel, 2... Coolant, 3... Reactor core, 4... Control rod, 121... Reactor protection device. Applicant's agent Patent attorney Takehiko Suzue Figure 1, 1 inch, Figure 2

Claims (2)

[Claims] (1) A nuclear reactor protection device that monitors the core flow rate of a nuclear reactor, detects that the core flow rate has decreased abnormally, and causes an emergency shutdown of the reactor. (2) A nuclear reactor according to claim 1, characterized in that the rate of decrease in core flow rate is calculated and the reactor is brought to an emergency shutdown when the rate of decrease in core flow rate exceeds a preset value. Protective device.