JPH05134082A - Method and system for measuring flow rate in reactor - Google Patents

Abstract

PURPOSE:To obtain a method and a system for measuring a flow rate in a reactor which can measure a core flow rate of the reactor highly precisely and stably on an on-line and real-time basis without increasing complicated in-pile instrumentation. CONSTITUTION:Local power range monitors 32A to 32D are provided in a core part 11 of a reactor and a computer 37 receiving LPRM signals as inputs from these local power range monitors 32A to 32D and subjecting them to a signal processing is provided. This computer 37 is equipped with a statistical processing circuit 38 which computes a rise speed A of a void from the LPRM signal, an analytical processing circuit 39 which computes a void fraction B from a reactor output and a void fraction estimation model based on this reactor output, an arithmetic processing circuit 40 which computes a liquid-layer rise speed C from the void rise speed A from the statistical processing circuit 38 and the void fraction B from the analytical processing circuit 39, and a flow rate computation processing circuit 41 which determines a sectional flow rate of a measured zone of section from the void fraction B, the void rise speed A and the liquid-layer rise speed C and determines a core flow rate Q by integrating each sectional flow rate, and thereby it measures the core flow rate Q passing through a core.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a technique for measuring a reactor core flow rate for measuring a core flow rate of a nuclear reactor, and particularly to a reactor flow rate for online and real-time measurement of a core flow rate of a boiling water reactor. And a measuring system therefor.

[0002]

2. Description of the Related Art In a boiling water reactor as a light water reactor, a core is housed in a reactor pressure vessel, and a plurality of jet pumps are arranged at intervals in the circumferential direction in a downcomer portion around the core. .. The jet pump is driven by the operation of the nuclear reactor recirculation system, and entrains the surrounding reactor coolant (reactor water) and guides it to the lower plenum of the core to ensure the required core flow rate.

To measure the core flow rate of a conventional boiling water reactor, a differential pressure measuring means is provided at the inlet and outlet of the jet pump, and the differential pressure at the jet pump inlet and outlet is measured by this differential pressure measuring means. The reactor coolant flow rate flowing through each jet pump was measured, and the core flow rate guided to the reactor core was measured.

[0004]

In the conventional boiling water reactor, the differential pressure at the jet pump inlet / outlet is measured to measure the core flow rate, so that the differential pressure is generated at the jet pump inlet / outlet port. In this case, the core flow rate can be measured accurately, but the type of reactor that does not have a jet pump like a natural circulation type reactor, or even if it is equipped with a jet pump, the jet in the natural circulation flow state When the differential pressure at the inlet and outlet of the pump was small and showed a small value, the flow rate measurement error of the core flow rate was large and the flow rate measurement could not be performed with high accuracy.

The present invention has been made in consideration of the above-mentioned circumstances, and it is possible to stably and accurately measure the core flow rate of a nuclear reactor in real time online without increasing the number of complicated in-core instrumentation devices. An object of the present invention is to provide a method of measuring the flow rate in a nuclear reactor and a measuring system therefor.

Another object of the present invention is to measure the core flow rate with high precision and accuracy by using the existing in-core instrumentation equipment even for an existing light water reactor, simplifying the flow rate measurement and improving the operability and It is possible to improve reliability.

[0007]

In order to solve the above-mentioned problems, the method for measuring the flow rate in a nuclear reactor according to the present invention, as set forth in claim 1, is a local power range for measuring the intensity of neutron flux. The LPRM signal from the monitor is input to the computer, and the LPRM signal is input to the computer.
In this method, the signal is processed by the correlation data between the reactor power and the flow pattern and the statistical mathematical method to measure the core flow rate through the core.

In order to solve the above-mentioned problems, the system for measuring the flow rate in a reactor according to the present invention has a local power range monitor installed in the core of the reactor, and an LPRM signal from the local power range monitor. A computer for inputting and processing the signal is provided, and this computer has a statistical processing circuit for calculating the rising rate of the void from the LPRM signal, and an analytical process for calculating the void ratio from the furnace output and the void ratio estimation model based on this furnace output. Circuit, an arithmetic processing circuit for calculating the liquid layer rising speed from the void rising speed of the statistical processing circuit and the void ratio of the analysis processing circuit, and the measurement cross-section area from the void ratio, the void rising speed and the liquid layer rising speed. It is equipped with a flow rate calculation processing circuit that calculates the cross-sectional flow rate and integrates each cross-sectional flow rate to obtain the core flow rate, and measures the core flow rate through the core. One in which the.

[0009]

The present invention positively utilizes the local power range monitor provided for reactor power instrumentation in a boiling water reactor, and inputs the LPRM signal from the local power range monitor to the computer to obtain the reactor output. The correlation data of the flow pattern is processed by a statistical mathematical method to obtain the void rate, the void (vapor phase) rising rate and the liquid layer rising rate, and the void rate, the void rising rate and the liquid layer rising rate are calculated. Then, the core flow rate rising in the reactor core can be measured with high accuracy in real time online.

The system for measuring the in-reactor flow rate according to the present invention can positively utilize the existing in-core instrumentation equipment provided for measuring the reactor power, and is complicated for measuring the core flow rate. There is no need to newly install an in-core instrumentation device.

Further, this method for measuring the flow rate in the nuclear reactor and its measuring system can accurately and accurately measure the core flow rate by processing the LPRM signal from the local power range monitor with a computer. Even if there is no jet pump in the core, or even if there is a jet pump, the inlet / outlet differential pressure is small and shows a very small value, the core flow rate can be measured accurately, simplifying flow rate measurement and improving operability and reliability. It is possible to improve.

[0012]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described below with reference to the accompanying drawings.

FIG. 1 is a schematic vertical sectional view showing a boiling water reactor as a light water reactor to which the present invention is applied. This boiling water reactor has a reactor pressure vessel 10 housed in a reactor containment vessel (not shown), and a reactor core 11 is housed in the reactor pressure vessel 10. The core 11 is surrounded by a core shroud 12, while an upper lattice plate 13 is provided in an upper part of the core shroud 12 and a core support plate 14 is provided under the core shroud 12. The core shroud 12 is covered by a shroud head 15.

An annular (sleeve-shaped) downcomer portion 17 is formed between the core shroud 12 and the inner peripheral wall of the reactor pressure vessel 10. Jet pumps 18 are circumferentially spaced from the downcomer portion 17. Multiple units are provided.
The jet pump 18 is configured to inject the coolant that is recirculated through the reactor recirculation system 19, and this coolant injection causes the reactor water around the downcomer unit 17 to be drawn in to cool the reactor coolant. It is guided to the lower plenum 20 below the core.

The reactor coolant guided to the lower core plenum 20 inverts and rises there, and is heated and boiled while passing through the core 11 to become a gas-liquid two-phase flow to form an upper core plenum 21. Sent to.

The gas-liquid two-phase flow guided to the core upper plenum 21 is guided to the steam / water separator 22 above the core, where it is separated into high-temperature water and steam. The separated steam is subsequently sent to the steam dryer 23 to be subjected to a drying action, becomes dry steam, and is sent from the main steam nozzle 24 to a steam turbine (not shown) to perform work. The steam that worked in the steam turbine is cooled in the condenser and becomes condensed water.

On the other hand, the high-temperature water separated by the steam separator 22 is mixed with the feed water sent from the condenser through the reactor condensate feed water system (not shown) and guided to the downcomer section 17. ..

Further, the core 11 of a boiling water reactor of 1100 MWe class is loaded with a large number of four fuel assemblies 25 as shown in FIG. 2 between the core support plate 14 and the upper lattice plate 13. A control rod 26 having a cross-shaped cross section is inserted between the four fuel assemblies 25 so that the control rod 26 can be inserted and removed. The control rod 26 is operated by a control rod drive mechanism (not shown).

In the core 11 of this boiling water reactor, the neutron source region monitor (SRM) detector 28 and the intermediate region monitor (I
The RM) detector 29 and the output region monitor (PRM) detector 30 are separately measured. Four SRM detectors 30 are provided to monitor the neutron flux in the neutron source 31 region, and the IRM detector 29 monitors the neutron flux in the intermediate region from the activated state to the full output.

Also, the PRM detector 30 is 1100 MWe
In the core 11 of the boiling water reactor of the class, for example, the PRM detectors 30 are provided at 43 places, and each PRM detector 30 is a local power range monitor (hereinafter referred to as LPRM) which is four independent neutron detectors as shown in FIG. ) 32A to 32D are arranged at intervals in the axial direction. Each P in the PRM detector 30
A movable in-core instrumentation (TIP) detector 33 is movably provided in the calibration conduit 34 for calibrating the PRMs 32A to 32D and measuring the neutron flux distribution in the axial direction of the core.

Thus, each detector 28, 29 and 30
The detection signal detected by, for example, each L of the PRM detector 30.
The LPRM signal, which is the detection signal from the PRMs 32A to 32D, is input from the signal cable 35 to the computer 37 shown in FIG. 4 through an output cable (not shown), and is processed by the computer 37 to measure the furnace output. Is becoming Since the reactor power changes due to fluctuations in the core flow rate as well as the control rod 26 being taken in and out, accurate measurement of the core flow rate is an important factor for watching the reactor power.

By the way, the system for measuring the in-reactor flow rate for measuring the core flow rate according to the present invention is based on the existing LPRM32.
It uses the LPRM signal from A to 32D and does not require addition of new in-core measuring equipment. Even without newly adding in-core measurement equipment, by processing the LPRM signal from each LPRM 32A to 32D by the computer 37, the core flow rate through the core 11 of the reactor can be accurately measured accurately in online real time. it can.

The computer 37 has the LPRMs 32A-3A.
The LPRM signal from 2D is input for signal processing. The statistical processing circuit 38 calculates the rising rate A of the void (vapor phase) from each input LPRM signal, and the void rate estimation model based on the reactor output. An analysis processing circuit 39 for calculating the ratio B, an arithmetic processing circuit 40 for calculating the liquid layer rising speed C from the void rising speed A and the void ratio B, a void rising speed A, a void ratio B, and a liquid layer rising The flow rate calculation circuit 41 calculates the cross-section flow rate D in the measurement cross-section area from the velocity C and integrates the cross-section flow rate D to obtain the core flow rate Q.

The computer 37 also has a simulation circuit 43 for the boiling water reactor core 11. The simulation circuit 43 incorporates known design data for simulation analysis data of the natural circulation BWR core and the forced circulation core. Known simulation analysis data include core data and LPRM data.

The simulation circuit 43 is connected to the statistical processing circuit 38 and the analysis processing circuit 39, and outputs known LPRM simulation data to the statistical processing circuit 38 and the analysis processing circuit 39.

On the other hand, the statistical processing circuit 38 receives LPRM signals from the LPRMs 32A to 32D of the respective PRM detectors 30, and outputs LPRM position information of the core 11, geometrical information around the instrumentation channel, pressure and temperature of the core 11, and the like. The void (vapor phase) rise rate A is obtained by statistically processing by a statistical mathematical method such as a Kalman filter based on the process information.

Each LPRM 32A-3 of the PRM detector 30
2D is installed for measuring the neutron flux intensity in the nuclear reactor, but if vapor voids 45 are generated in the reactor core 11, the generated vapor voids 45 affect the neutron flux intensity, and each LPRM 32A ~ LPRM detected at ~ 32D
As shown in FIG. 5, the signal undergoes a change in neutron flux intensity, which causes a signal change. Therefore, each LP of the PRM detector 30
By processing the LPRM signals detected by the RMs 32A to 32D and measuring the signal change and the time delay thereof, the passage of the void can be measured. The statistical processing circuit 38 statistically processes a signal change of the LPRM signal from each LPRM 32A to 32D and its time delay, that is, the time delay of the neutron flux intensity by a statistical mathematical method such as a Kalman filter to obtain the void rising rate. There is.

Further, the analysis processing circuit 39 uses the LPRMs 3
An analysis calculation circuit that inputs LPRM signals from 2A to 32D to analyze core nuclear characteristic information and core process (temperature / pressure) information, and obtains quality which is reactor power change, core reactivity change, and steam weight ratio 47 and an analysis estimation circuit 48 for estimating the void ratio B by estimating the void ratio statistically by the statistical mathematical method on the data calculated and analyzed by the analysis calculation circuit 47. From the calculated quality using the known void-quality correlation equation 49 and the one obtained using the simulation circuit from the reactor power and reactivity changes, the statistical estimation method such as Kalman filter is used in the analysis estimation circuit. The void rate B is estimated and measured.

Thus, the void rising speed A obtained by the statistical processing circuit 38 and the void rate B found by the analysis processing circuit 39 are inputted to the arithmetic processing circuit 40, and the inputted void rising rate A and the whid rate are inputted. From B and B, the liquid layer rise rate C is calculated and estimated by taking into account the flow pattern flow correlation data based on the known drift flux model and the drift velocity. Alternatively, the difference in rising speed between the liquid layer and the gas layer (void) can be obtained from the slip ratio, and the rising speed of the liquid phase can be estimated from the slip ratio.

In this way, each LPRM 32A-32
The LPRM signal from D is statistically processed by the statistical processing circuit 38 to obtain the whid rise rate A, and the LPRM signal is analytically processed by the analysis processing circuit 39 to obtain the void rate B. The liquid layer rising speed C is obtained by arithmetically processing the A and the void ratio B in the arithmetic processing line 40 in consideration of the correlation data of the flow pattern, the drift speed, and the like.

The void rising speed A, the void ratio B, and the liquid layer rising speed C are calculated by the flow rate calculation processing circuit 41.
Is input to and processed. In the flow rate calculation processing circuit 41, the cross-sectional flow rate Qi for each measurement region of each PRM detector 30 is obtained. This cross-sectional flow rate Qi is

[0032]

[Equation 1]

The cross-sectional flow rate Q obtained by
The core flow rate Q is obtained from i by the following equation.

[0034]

[Equation 2]

The void density and liquid layer density are known physical property data obtained from the process information (temperature, pressure, etc.) of the core.

In this way, each LPRM 32A-32
By processing the LPRM signal from D by the computer 37 by a statistical mathematical method, the core flow rate can be accurately measured in real time online.

Next, the principle of measuring the core flow rate using the system for measuring the flow rate in the reactor is shown in FIG. 6, and the block diagram for measuring the core flow rate is shown in FIG.

This measurement system uses L from LPRMs 32A to 32D built in the core of a boiling water reactor.
Using the PRM signal, the LPRM signal is subjected to signal processing by a computer 37 that has a built-in statistical mathematical method and a correlation between the reactor output and the flow mode in advance to obtain the core flow rate. This measurement system is for each LPRM32A
32D is input to the computer 37, the flow rate is estimated by a plurality of flow rate estimation models 50 and 51, and the core flow rate is estimated from each estimated flow rate by the optimization theory that minimizes the error by a statistical mathematical method. However, it is designed to measure online.

This measuring system is concretely signal-processed by the block diagram schematically shown in FIG.

Each LPRM 32A loaded in the core 11
With respect to the LPRM signal from 32D, the void rising speed A for each section is obtained by the statistical mathematical method by the block 1 which is the statistical processing circuit 38. Block 1 is provided with a Kalman filter incorporating, for example, an ascending model of steam void slag.

Further, each LPRM signal is analyzed by the analysis processing circuit 39.
The void rate B for each section is estimated by the block 2 which is a statistical mathematical method. The block 2 is provided with, for example, a void fraction estimation model Kalman filter based on neutron flux characteristics or a statistical estimation circuit as the analysis estimation circuit 48.

Here, the Kalman filter 53 is
LPRMs obtained as shown in FIG. 8 and obtained sequentially
A physical quantity estimated value in which the reactor measurement data such as a signal is corrected by the measurement process model 55 incorporating the measurement accuracy information,
From the process control information and process amount variation information related to reactor operation / control parameters to the plant physical model 56
And the physical quantity estimated value estimated by using the Kalman gain K so as to minimize statistically and mathematically, and add the physical quantity estimated value with high accuracy (for example, void rate or void rising speed). ). The Kalman gain K is a value that is sequentially calculated from the measurement accuracy information and the deviation between the filter estimated value and the measurement data.

Further, in the above Kalman filter 53, the measurement accuracy information is information obtained from data and experience on the specifications and performance of the LPRM measurement sensor, and the process control information is obtained from knowledge about the plant design and operating state of the reactor. Further, the process amount variation information is information acquired from knowledge such as specifications of plant equipment of the reactor, operation modes, and actual measurement. The Kalman filter 53 is used in the statistical processing circuit 38 and the analysis / estimation circuit 48 of the computer 37 to accurately estimate and measure the physical quantities of the void rising speed A and the void rate B. You may use the optimization theory which makes the minimum error by a statistical mathematical method other than the Kalman filter 53.

The void ascending velocity A and the void ratio B thus obtained are input to the block 3 which is the arithmetic processing circuit 40, and the liquid is calculated from the known drift flux model based on the void distribution constant and the drift velocity. The layer rising speed C is estimated and calculated. The block 4 is provided with a void distribution constant, a drift velocity, a void-quality correlation equation, and known data of a flow pattern in advance as a physical property value table, and is included in the simulation circuit 43 of FIG. 4, for example.

Further, the flow rate calculation processing circuit 41 considers the known void density and liquid phase density from the void rise rate A, the void rate B, and the liquid layer rise rate C estimated and obtained in blocks 1 to 3. The core flow rate Q is accurately estimated and obtained by the block 5.

[0046]

As described above, in the present invention, the local power range monitor provided for the reactor power instrumentation of the boiling water reactor is positively used, and the LPRM signal from the local power range monitor is used. Is input to the computer and the correlation data between the furnace output and the flow pattern is processed by a statistical mathematical method to obtain the void rate, the void rise rate, and the liquid layer rise rate, and the void rate, the void rise rate, and the liquid layer The rising velocity can be calculated to accurately measure the core flow rate in the reactor core in online real time with high accuracy.

Since the system for measuring the in-reactor flow rate of the present invention positively utilizes the existing in-reactor instrumentation equipment provided for reactor power instrumentation, a complicated in-reactor instrumentation device is newly added. No need.

Further, according to the present invention, since the LPRM signal from the local power range monitor is processed by a computer so that the core flow rate can be accurately measured accurately, the jet pump can be used like a natural circulation core. Even if there is no jet pump, or if there is a jet pump and the differential pressure at the pump inlet and outlet is small and takes a very small value, the core flow rate can be accurately measured, and flow rate measurement can be simplified and reliability can be improved. ..

[Brief description of drawings]

FIG. 1 is a vertical cross-sectional view of a boiling water reactor equipped with a system for measuring a flow rate in a reactor according to the present invention.

FIG. 2 is a plan view showing a core structure of the boiling water reactor.

FIG. 3 is a vertical cross-sectional view showing a PRM detector loaded in the core of a boiling water reactor.

FIG. 4 is a diagram showing an embodiment of a system for measuring a flow rate in a nuclear reactor according to the present invention.

FIG. 5 is a graph showing a relationship between an output signal from each LPRM of the PRM detector and void passage.

FIG. 6 is a schematic diagram showing a measurement principle of measuring a flow rate in a reactor by a statistical mathematical method.

FIG. 7 is a block diagram showing a measuring method used in the system for measuring the flow rate in a nuclear reactor according to the present invention.

FIG. 8 is a diagram illustrating a Kalman filter included in the block diagram of FIG. 7.

[Explanation of Codes] 10 Reactor Pressure Vessel 11 Core 12 Core Shroud 15 Shroud Head 17 Downcomer 18 Jet Pump 20 Lower Core Plenum 21 Upper Core Plenum 22 Steam Separator 23 Steam Dryer 25 Fuel Assembly 26 Control Rod 30 PRM Detector (output area monitor detector) 32A to 32D LPRM (local power range monitor) 37 Computer 38 Statistical processing circuit 39 Analysis processing circuit 40 Calculation processing circuit 41 Flow rate calculation processing circuit 43 Simulation circuit 45 Steam void 47 Analysis calculation circuit 48 Analysis Estimating circuit 53 Kalman filter A Void rise rate B Void rate C Liquid layer rise rate D Cross sectional flow rate in measured cross sectional area (Qi) Q Core flow rate

Claims (2)

[Claims] 1. A LPRM signal from a local power range monitor (LPRM) for measuring the intensity of a neutron flux is input to a computer, and the LPR signal is input to the computer.
A method for measuring a flow rate in a reactor, which comprises subjecting an M signal to signal processing by a correlation data between a reactor output and a flow mode and a statistical mathematical method to measure a core flow rate through a core. 2. A local power range monitor (LPRM) is installed in a core portion of a nuclear reactor, and a computer for inputting an LPRM signal from the local power range monitor to perform signal processing is provided. The computer is a void from the LPRM signal. Statistical processing circuit for calculating the rising rate of, the furnace output and an analysis processing circuit for calculating the void rate from the void rate estimation model based on this furnace output, the void rising rate of the statistical processing circuit and the void rate of the analysis processing circuit A calculation processing circuit for calculating the liquid layer rising speed from the flow rate calculation processing for calculating the cross-sectional flow rate of the measurement cross-sectional area from the void rate, the void rising speed, and the liquid layer rising speed, and calculating the core flow rate by integrating each cross-sectional flow rate And a circuit for measuring the core flow rate passing through the core.