JPH02143194A - Flow rate measuring method in core - Google Patents

Abstract

PURPOSE:To perform operation monitoring, protection, etc., with high accuracy and to improve the reliability of nuclear reactor operation by calibrating and correcting the result of a core support plate part differential pressure measuring method periodically and continuously with the measured value of another different measuring method. CONSTITUTION:In this measuring method, a differential pressure signal from a support part differential pressure transmitter 7 varies under the influence of a void quantity generated according to the output of a nuclear reactor, so the differential pressure variation is corrected at all times with core flow rate- core support part differential pressure characteristics by an output correction signal inputted through a mean output area circuit 15 from a neutron flux signal. Further, a support plate flow rate arithmetic part 8 performs calibration periodically with the core flow rate signal of an internal pump exit/entrance part differential pressure measuring method from the exit/entrance part flow rate arithmetic part 12 and also performs calibration periodically with the core flow rate signal of a thermal balancing measuring method from a thermal balance flow rate computing element 16 to compute and output a core flow rate signal which is improved in accuracy.

Description

DETAILED DESCRIPTION OF THE INVENTION [Object of the Invention] (Industrial Application Field) The present invention relates to a method for measuring the core flow rate of a nuclear reactor.

(Prior technology) The core flow rate of a boiling water reactor is an extremely important parameter from the standpoint of reactor reactivity control. Conventionally, the core flow rate was created using a primary coolant recirculation drive flow using a recirculation pump and a jet pump. is controlling this. This jet pump is a static flow rate element, and an accurate method for measuring the core flow rate is obtained by measuring the differential pressure in the diphasor section and determining each discharge coefficient in a factory test. In addition to this, we are also measuring the differential pressure at the core support plate, but we cannot expect it to be accurate.

Therefore, in order to increase the reliability of the measurement results, monitoring can be performed using a separate measurement method.

It is also being put into practical use recently. In a boiling water reactor plant that uses an internal pump that does not use a recirculation pump installed outside the reactor, the system includes differential pressure measurement at the internal pump deck instead of the conventional jet pump duphasor differential pressure measurement. A method of measuring the differential pressure between the inlet and outlet of a pump is known. However, although this method of measuring the differential pressure flow rate at the inlet and outlet of the internal pump is highly accurate, since the internal pump is a dynamic device, it requires input conditions for multiple internal pump rotational speeds. Individual differences and the influence of the flow state of the coolant inside the reactor must be taken into account, and this requires complex calculations, resulting in difficulties in response, and therefore higher accuracy than conventional methods. It was difficult to obtain.

Furthermore, during partial operation of a plurality of internal pumps, backflow occurs through the stopped internal pumps, making it difficult to obtain a more accurate core flow rate. The method for measuring the core support plate differential pressure is as follows:
As shown in the characteristic diagram in Figure 3, the relationship between the differential pressure at the core support plate and the core flow rate is such that the amount of voids inside the reactor changes depending on the reactor power, which changes the core pressure drop and the aging of the core. Because there is a pressure drop change, this affects the differential pressure measurement, making it impossible to obtain highly accurate core flow measurement. For this reason, it was necessary to calibrate and use the measurement value obtained by the above-mentioned internal pump inlet and outlet differential pressure measurement method. However, once this measurement method has been calibrated, it does not require particularly complicated calculations and has relatively good responsiveness. Since it has been experimentally known that the flow velocity is approximately uniform at the plate position, the advantage is that no calculation processing is required.

(Problem M to be solved by the invention) In a boiling water reactor that generally employs an internal pump, the method for measuring the differential pressure at the inlet and outlet of the internal pump and the method for measuring the differential pressure at the core support plate are mainly used on the plant system. In response to demands 1. For example, high accuracy is required for plant performance calculations.

The differential pressure measurement method at the inlet and outlet of the internal pump was used, and the responsive differential pressure measurement method at the core support plate was used for the safety protection system function, which caused problems in the complexity of switching between selections and the reliability of the operation results. was there.

The present invention has been made in view of the above, and its purpose is to provide a core support plate differential pressure measurement method with good responsiveness when measuring a core flow rate.

The purpose of the present invention is to provide a method for measuring core flow rate with higher accuracy and improved reliability by performing calibration from the results of multiple other measurement methods that provide high accuracy.

(Means for Solving the Problem) Calibration is performed using the core support plate differential pressure measuring means, the rostral internal pump inlet/outlet differential pressure measuring means, and the flow rate measuring means based on the inlet temperature of the lower part of the reactor based on the thermal balance of the reactor. The system is equipped with a correction means using a neutron flux detector installed in the neutron flux detector.

(Function) The core flow rate measured by the core support plate differential pressure measurement method, which has good responsiveness but is easily affected by the amount of voids in the coolant, is compared to the core flow rate measurement method, which has high measurement accuracy but is easily affected by the amount of voids in the coolant. In addition to periodically calibrating the flow rate measured by measuring the differential pressure at the entrance and exit of the internal pump, which requires consideration of the conditions, and the flow rate measured by calculating the thermal balance of the reactor, the influence of the amount of voids in the coolant detected by a neutron flux detector The core flow rate can be measured with high precision and responsiveness by correcting the

(Example) An embodiment of the present invention will be described with reference to the same work.

FIG. 1 is an overall configuration diagram, in which a plurality of internal pumps 3 installed around the lower part of the reactor core 2 in the reactor pressure vessel 1 move the coolant 4 in the reactor pressure vessel 1 from the lower part of the reactor core 2 to the upper part. A flow is created towards. Further, there is a core support plate 5 at the bottom of the core 2.

A support plate differential pressure transmitter 7 is provided which is connected via instrumentation pipes 6a and 6b connected to nozzles that are opened at the top and bottom of this, and this output differential pressure signal is used to calculate the support plate flow rate. input into device 8. Further, instrumentation piping 10a, 10b is connected to a nozzle opened at the pump deck 9 section of the internal pump 3 or the inlet/outlet section of the internal pump 3.
An inlet/outlet differential pressure transmitter 11 is provided, and the output differential pressure signal is input to an inlet/outlet flow rate calculator 12. Furthermore, a neutron flux detector I3 is installed in the core 2, and a temperature detector 14 for the coolant 4 is installed at the lower inlet of the core 2.
The neutron flux signal from the neutron flux detector I3 is input to the support plate flow rate calculator 8 via the average output area circuit 15. The temperature signal from the temperature detector 14 is sent to the thermal equilibrium flow rate calculator 1.
6, this heat balance flow rate calculator 16 calculates the energy heat balance to the reactor by making the temperature at the inlet of the downcomer 16 and the core support plate 5 represented by the temperature detector 14. , calculate the core flow rate value using the heat balance tM (heat balance) 81g determination method for this reactor, and output this output to the support plate flow rate calculator 8. In addition, the inlet/outlet flow rate calculator 12 obtains information from the inlet/outlet differential pressure signal from the inlet/outlet of the internal pump 3, the rotation speed information of each internal pump 3 (not shown), and the test results of each internal pump 3 at the factory. From the obtained Q=H characteristic information, the core flow rate is calculated by the method of measuring the differential pressure at the internal pump inlet and outlet, and is output to the support plate flow rate calculator 8, and is also used as an indicator or recorder for other purposes as necessary. Output to 17. The support plate flow rate calculator 8 calculates the core flow rate using the core support plate differential pressure measurement method, outputs the result to the indicator or recorder 18 according to system requirements, and controls the flow rate. It is configured to serve as an interlock signal for signals and other equipment.

Next, the effect of the above configuration will be explained. Figure 2 is an operational flow diagram, in which the support plate flow rate calculator 8 calculates the core flow rate with good responsiveness from the core support plate differential pressure (ΔP) using the core support plate differential pressure measurement method. In this method, the differential pressure signal from the support plate differential pressure transmitter 7 changes due to the effect of the amount of voids generated depending on the output of the reactor.
From the output correction signal (P) inputted from the neutron flux signal through the average power region circuit 15, the differential pressure change is determined in advance by analysis or reactor start-up test. Constant correction is made depending on the characteristics.

The support plate flow rate calculator 8 is periodically calibrated using the core flow rate signal from the inlet/outlet flow rate calculator 12 using the internal pump inlet/outlet differential pressure measurement method, and furthermore, the support plate flow rate calculator 8 is calibrated using the core flow rate signal from the inlet/outlet flow rate calculator 12 using the internal pump inlet/outlet differential pressure measuring method. A core flow Q signal (W) whose accuracy is improved by periodically calibrating the core flow signal (C) determined by the measurement method is calculated and output. This core miscarriage signal with a small error is used not only for core flow measurement control as well as for example, the reference flow fJcl! for neutron flux monitoring, as in the past.
1g constant, scram function for rapid decrease in core flow rate, etc.
The information is transmitted to various interlocks and monitoring devices in the protection system, improving the reliability of reactor operation.

Here, a rough calculation of the core flow of 5 kV/t, which is determined by the heat balance (heat balance) of 31 g, can be performed using the following equation.

Here, ho; hf; hfg; fcu; cr er p h.

W c u v hcuw; Wry; hfw; Heartthrob Roentalbi, Saturated water enthalpy.

Saturated steam enthalpy, saturated steam flow rate entering the downcomer, rtIJ J/4 dynamic system Calano flow rate, enthalpy from the control rod drive system, internal pump purge flow rate, internal pump purge intarbi, inflow flow rate from the reactor purification system .

Inflow enthalpy from the reactor purification system.

Reactor feed water flow rate, Reactor feedwater enthalpy.

Q+-; Thermal energy C2 lost in the downcomer part
; Conversion constant (860Kcal/kWh).

Ql,; Thermal energy obtained by the internal pump.

Therefore, for the measurement error (100%) that occurs when only the core support plate differential pressure measurement method is used, the effect of error correction using the internal pump inlet and outlet differential pressure measurement method is as shown in equation (2) below.

Σ, =E777...■ Here, ea 7 Error in the differential pressure measurement H at the pump inlet and outlet,
ec; Error in core support plate differential pressure measurement.

Furthermore, the effect of error correction using the flow rate measurement method based on thermal equilibrium is expressed by the following equation 0.

Σ2 = fi/4(ed"+eb") -e(,,"
...■ Here, cb; Error measured from thermal equilibrium.

From the above, now e &= 6 b: e. Assuming that, according to the present invention, the reduction can be improved to Σ1/Σ,=1/3/2=87%.

〔Effect of the invention〕

As described above, according to the present invention, when measuring the core flow rate of a nuclear reactor that employs an internal pump, the results of the core support plate differential pressure measurement method are periodically and continuously calibrated and corrected using the values measured by other different measurement methods. By doing so, even under various operating conditions of multiple internal pumps, the response is good and accurate core flow rate can be obtained, which makes it possible to control the reactor output.

Operation monitoring, protection, etc. are performed with high precision, and the efficiency of reactor operation is improved.
This has the effect of improving f reliability.

[Brief explanation of the drawing]

FIG. 1 is an overall configuration diagram of an embodiment of the present invention, FIG. 2 is an operational flow diagram according to the present invention, and FIG. 3 is a characteristic diagram of core flow rate and core support plate differential pressure with respect to changes in reactor output. 2... Core, 3... Internal pump, 5... Core support plate. 7... Core support plate differential pressure transmitter, 8... Support plate flow rate calculator. 11... Inlet/outlet differential pressure transmitter. I2... Inlet flow rate calculator, 13... Neutron flux detector, 15... Average output area circuit, 16... Thermal equilibrium flow rate calculator. 17, 18...Recorder. Technique 4...Temperature detector.

Claims (1)

[Claims] In a boiling water reactor in which an internal pump is installed in the reactor to control the core flow rate, the flow rate is measured by the differential pressure at the entrance and exit of the internal pump, compared to the flow rate measurement method using the differential pressure at the core support plate. Calibration is performed using the measured value of the core flow rate calculated by the flow rate measurement method based on the thermal equilibrium of the reactor from the lower core inlet temperature, and the change in core pressure drop due to the effect of void volume is corrected using the in-reactor neutron flux signal. A core flow rate measurement method characterized by: