JPH06130177A - Reactor monitoring device - Google Patents

Abstract

(57) [Summary] [Purpose] While directly calculating the fluid flow rate from the outside of the pipe, calculating nuclear and thermal process data, calibrating the existing detector, and monitoring for abnormalities, (EN) Provided is a reactor monitoring device capable of predicting signs of deterioration, failure, etc., replacing, and clarifying repair times. [Structure] A first invention is a radiation measuring means for continuously measuring a dose rate by a radiation detector 14 installed near a main steam pipe and the like, and a correlation between the dose rate and steam flow rate, nuclear instrumentation value, and heat output. Calculate the nuclear instrumentation value, heat output or main steam flow rate of the reactor from the characteristics and monitor the signs of abnormalities in the existing detector by comparing it with the process data from the existing main steam flow meter 10 and in-core neutron detector 11. It consists of arithmetic processing means 16. The second invention is a radiation measuring means for continuously measuring the dose rate detected by a plurality of radiation detectors 27, 28 which are installed in the vicinity of the main steam pipe and the like in a time series, and the dose rate of the measurement target is pulsed. The hydrogen injecting device 26, the flow rate from the transition time of the dose rate change and the separation distance, the flow rate is calculated from the flow rate and the pipe cross-sectional area, etc., and the existing flow meter 10 is calibrated, and deterioration and failure are predicted. It is characterized by comprising an arithmetic processing means 29.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to measurement of heat output, main steam flow rate, feed water flow rate, etc. of a nuclear reactor, and in particular process data is calculated by radiation measurement, and existing main steam flow rate, feed water flow rate, etc. The present invention relates to a reactor monitoring device that can calibrate a detector and periodically inspect it to monitor for abnormal signs such as deterioration and malfunctions, collect data, and predict repair / replacement times.

[0002]

2. Description of the Related Art Conventionally, in a boiling water reactor, as shown in the schematic system diagram of FIG. 8, a reactor containment vessel 2 is installed in a nuclear building 1, and the inside of this reactor containment vessel 2 is installed. A reactor pressure vessel 4 accommodating the core 3 is housed in. The coolant heated in the reactor core 3 in the reactor pressure vessel 4 becomes superheated steam, flows into the turbine 6 through the main steam pipe 5, and the steam causes the turbine 6 to rotate and the generator 7
Drive to generate electricity.

The steam that has finished its work in the turbine 6 flows into the condenser 8, is cooled by seawater and becomes condensed water, passes through the water supply pipe 9 and returns again into the reactor pressure vessel 4, and is heated in the core 3. To be done. A main steam flow meter 10 is attached to the main steam pipe 5. Further, an in-core neutron detector 11 for measuring heat output is provided in the core 3, and is arranged in a space sandwiched by fuel assemblies (not shown) in the core 3 (one per 16 bundles), The neutron flux in the reactor is directly measured.

Further, the reactor pressure vessel 4 is provided with a coolant recirculation pipe 12 and a purification system pipe 13 between the coolant recirculation pipe 12 and the water supply pipe 9. Here, main flowmeters such as the main steam flowmeter 10 used in the reactor process system are provided with a throttle mechanism in the fluid flow path and a throttle flowmeter (orifice) for measuring the differential pressure before and after , Nozzle type (nozzle Venturi tube), etc., all of which are arranged in the fluid flow path in the pipe.

The neutron detector 11 in the reactor is mainly composed of ²³⁵
It is a nuclear fission ionization chamber with a U, with the operation of the reactor ²³⁵
Since the sensitivity decreases due to the splitting of U, a movable in-core detector (not shown) is regularly inserted from outside the reactor during the period of use to calibrate the in-core neutron detector 11 and the period of use is long. It is replaced at the time of periodic inspection of the reactor.

On the other hand, plant data such as reactor heat output, main steam flow rate, feed water flow rate, etc. are calculated by a process computer based on data of main process systems in addition to the direct values obtained from the measurement system. There is also plant data,
The heat output is calculated from the feed water flow rate and main steam flow rate.

[0007]

[Problems to be Solved by the Invention]
Monitoring of thermal data is an important measurement item that is indispensable for safe operation and control of a nuclear reactor plant, and among the flow rate measurements in a nuclear reactor, the most important one is the coolant in the main loop of the primary and secondary systems. And steam flow measurement. Although there are various methods for measuring the flow rate, the main flow meter that is usually used in the reactor process system is a throttle flow meter that measures the differential pressure before and after it by providing a throttle mechanism in the fluid flow path. . The main steam system flow meter used here is a nozzle Venturi tube in the throttle flow meter.

However, in this conventional method, the flow velocity and the flow rate of the fluid are not directly measured but are indirectly determined, so that an error is likely to occur. Therefore, in order to measure the vapor flow rate within the allowable accuracy, the flow rate and the flow rate must be upstream of the flow rate detecting element. In addition to having to provide a straight pipe part with a certain length downstream, it is necessary to perform processing such as drilling holes in the pipe because the orifice is inserted in the pipe and the pressure gauge is attached in the method using differential pressure measurement. There is a possibility that coolant leaks from this processing part,
And there is a problem such as a failure of the flow rate detecting element.

Further, since the inside of the pipe of the primary system is subject to inspection and replacement of the flow meter and the like, there is a problem that radioactive contamination of the flow meter and the like and exposure to workers increase. Conventionally, the heat output of a nuclear reactor has been obtained by calculation from the feed water flow rate and the main steam flow rate. However, there is an in-core neutron instrumentation as another measuring device, and a small in-core neutron detection can be performed directly in the core 3. vessel
11 is placed and the neutron flux is measured and converted into heat output.

Also in this in-core neutron detector 11, the environment in which the detector is installed is high temperature and high pressure, and due to the high radiation dose field, the secular change in the sensitivity of the in-core neutron detector 11 poses a problem. Therefore, a movable in-core neutron detector (not shown) for the purpose of periodically performing the calibration work of the in-core neutron detector 11 was installed outside the reactor, and was arranged in the core 3 as necessary. The neutron detector 11 in the reactor is inserted up to the position for calibration.

The measurement systems such as the conventional main steam flow meter 10 and the in-core neutron detector 11 cannot be directly calibrated after the installation, and the detector is directly exposed to the object to be measured and placed in a harsh environment. Since this may cause deterioration, breakdown, and accidents, it is an issue to develop a device that performs performance inspection after installation and easily monitors (confirms) abnormalities such as deterioration and breakdown. In addition, it is easy to directly calibrate the main steam flow meter and feed water flow meter of the primary system as they are already installed, and predict the signs of deterioration, failure, etc. of the relevant flow meter and clarify the timing of replacement and repair. There has been a demand for a device that can.

The first object of the present invention is to measure radiation contained in a fluid from the outside of the main steam pipe or the water supply pipe during the operation of the reactor to directly calculate the fluid flow rate, and Calculate process data such as physical and thermal nuclear instrumentation values and monitor the existing process data for signs of abnormalities in detectors and grids. Secondly, the flow rate of the fluid can be directly calculated to easily directly calibrate the existing main steam flow meter and feed water flow meter, and at the same time predict the deterioration and failure of the flow meter and replace it. -To provide a reactor monitoring device that can clarify the timing of repairs.

[0013]

A first aspect of the present invention is a radiation detector installed in the vicinity of piping such as a main steam system or a primary cooling system of a nuclear reactor and a radiation dose rate detected by the radiation detector. It consists of radiation measuring means for measurement and arithmetic processing means for calculating the nuclear instrumentation value, heat output or main steam flow rate of the reactor from the correlation characteristics of the radiation dose rate and the main steam flow rate, nuclear instrumentation value, heat output. Is characterized by. Further, in the arithmetic processing means, it is possible to monitor the sign of abnormality of the existing detector by comparing the radiation dose rate detected by the radiation measuring means with the nuclear instrumentation value or the process data.

A second aspect of the present invention is to provide a plurality of radiation detectors which are installed in the vicinity of pipes such as a main steam system and a primary cooling system of a nuclear reactor at a predetermined distance, and a radiation dose detected by the plurality of radiation detectors. From the radiation measuring means for continuously measuring the rate in time series, the hydrogen injection device for changing the radiation dose rate to be measured in a pulsed manner, the transition time of the radiation dose rate change and the predetermined separation distance of the radiation measuring means. It is characterized in that it comprises arithmetic processing means for calculating the flow rate of the fluid in the main steam system, the primary cooling system, etc. from the flow rate and the pipe cross-sectional area, etc., and for performing comparative calibration with the measured value of the existing flow meter. . In addition, the arithmetic processing means periodically monitors and compares the measured values of the existing flow meter to predict the slight symptoms such as deterioration and failure of the existing flow meter.

[0015]

In the first aspect of the invention, the radiation detector installed near the pipe of the main steam system or the like detects the radiation dose rate from the nuclide contained in the fluid in the pipe and continuously measures it in the radiation measuring means. The calculation processing means directly calculates the nuclear instrumentation value, the heat output, the main steam flow rate, etc. of the nuclear reactor from the correlation characteristics between the radiation dose rate and the process data such as the main steam flow rate, the nuclear instrumentation value, the heat output. Further, in the arithmetic processing means, the abnormality of the existing detector and its symptom can be monitored from the time series comparison of the radiation dose rate detected by the radiation measuring means and the process data such as the main steam flow rate and the heat output.

According to a second aspect of the present invention, a plurality of radiation detectors are installed in the vicinity of a pipe such as a main steam system at a predetermined distance from each other, and nuclides contained in a fluid in the pipe detected by the plurality of radiation detectors. The radiation dose rate is continuously measured in time series by the radiation measuring means, and the hydrogen injection device injects hydrogen into the fluid so that the radiation dose rate of the nuclide to be measured is changed in a pulsed manner, and the arithmetic processing is performed. The flow velocity of the fluid is calculated from the transition time at which the radiation dose rate changes by the means and the distance between the radiation detectors.

Further, the flow rate of the fluid is calculated from the flow velocity and the cross-sectional area of the pipe, and the measured value of the existing flow meter is compared and calibrated. Further, in the arithmetic processing means, the measured flow rate value of the fluid and the flow rate value of the existing flow meter are monitored and compared to predict the abnormality such as the deterioration and the failure of the existing detector.

[0018]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to the drawings. It should be noted that the same components as those of the above-described conventional technique are denoted by the same reference numerals and detailed description thereof will be omitted. 1st invention WHEREIN: As shown in the schematic system block diagram of FIG. 1, the reactor containment vessel 2 is installed in the reactor building 1, and the reactor core 3 was accommodated in this reactor containment vessel 2. Reactor pressure vessel 4
Is stored. The coolant heated by the core 3 in the reactor pressure vessel 4 becomes superheated steam, flows through the main steam pipe 5, and flows into the turbine 6. Turbine 6 with this steam
Is rotated and drives the generator 7 to generate electricity.

The steam that has finished its work in the turbine 6 is returned to the condenser 8
To be condensed by being cooled by seawater, returned to the inside of the reactor pressure vessel 4 through the water supply pipe 9 and heated in the core 3. An existing main steam flow meter 10 is attached to the main steam pipe 5, and an existing in-core neutron detector 11 is also installed in the core 3 to directly measure the neutron flux in the reactor. Is going.

Further, the reactor pressure vessel 4 is provided with a coolant recirculation pipe 12, and a purifying system pipe 13 is provided between the recirculation pipe 12 and the water supply pipe 9, and the reactor core 3 Part of the boiling water component (primary coolant) of the coolant heated in (1) is purified from the recirculation pipe 12 through the purification system pipe 13 and returned to the water supply pipe 9.

As the measurement system, the existing main steam pipe 5
In addition to the main steam flow meter 10 which measures the differential pressure with a venturi type flow restrictor attached to the reactor, and the in-core neutron detector 11 which measures the neutron flux arranged in the core 3, it is a radiation measuring means. ¹⁶ N installed near the main steam pipe 5 and
There is a radiation detector 14 whose measurement target is ¹⁵ C. The radiation detector 14 is a plastic scintillation detector or the like and continuously measures the dose rate or count rate.

In the main steam system during normal operation, ¹⁶ N, ¹⁵ C and ¹⁷ N which are the radiation sources of the radiation detector 14 and the reactor water
¹³ N, etc. are generated in the core and have a short half-life, and they are transferred to the steam system and reactor water (purification system) depending on their chemical forms. In addition,
The radiation detector 14 can be installed near the purification system pipe 13, and the output of the radiation detector 14 is configured to be transmitted to the arithmetic processing means 16 together with the output of the plant data output device 15.

The block diagram of FIG. 3 shows the details of the radiation measuring means and the arithmetic processing means 16. The radiation measuring means is a radiation detector 14 installed in the main steam pipe 5 or the purification system pipe 13, Connected via a signal cable 19 to a power supply 17 for applying voltage and a dose rate data memory 18,
It is formed by a dose rate data memory 18 for storing measured data and an optical transmission module 20, and the optical transmission module 20 is connected to the arithmetic processing means 16 by an optical cable 21.

Data from the plant data output device 15 is input to the arithmetic processing means 16, and a data output device 22 such as a graphic printer is connected to the arithmetic processing means 16.
Steam flow rate-piping dose (thermal output-piping dose) correlation diagram, as shown in the correlation diagram, calibrated various correlation characteristics based on actual measurement data (plant data for radiation dose rate at each measurement point). Built-in, in addition to calculating the fluid flow rate and heat output using this correlation characteristic,
This calculated plant data and the existing main steam flow meter 10
Also, the presence of abnormalities in existing detectors is monitored by comparison with plant data measured by the in-core neutron detector 11 and the like.

Next, the operation of the above configuration will be described. Radiation detector 14 installed near the main steam pipe 5
Is ¹⁶ N contained in the main steam flowing through the main steam pipe 5 and
The radiation dose rate or count rate of ¹⁵ C is continuously measured, and the measured data is stored in real time in the dose rate data memory 18 as a digital value, and the data is processed through the optical transmission module 20 and the optical cable 21. Transfer to.

From the input data, the arithmetic processing means 16 pays attention to the fact that there is a proportional relationship between the steam flow rate and the pipe dose (heat output-pipe dose) as shown in FIG. The main steam flow rate in the main steam pipe 5 is calculated using this. Moreover, the soundness of the existing main steam flow meter 10 and in-reactor neutron detector 11 can be measured by processing the radiation dose rate measurement data measured in succession in time series and comparing with the plant data measured by the existing detector. Can be confirmed.

That is, in the arithmetic processing means 16, the radiation data and the plant operation data (main steam flow rate, heat output, etc.) at the time of measuring the dose rate are displayed in time series and compared, so that the characteristics shown in FIG. As shown in the curve diagram, the soundness can be confirmed by comparing the radiation dose 23 in the main steam pipe 5, the main steam flow rate 24, and the change in plant data such as the heat output 25 with time by the data output device 22.

If there is a change in the plant data output due to an abnormality in the existing detector for the plant data or the signal processing system, or due to the symptom thereof, the radiation data between the radiation detector 14 and the radiation data The change tendency over time does not correspond, and the abnormality can be predicted at a glance.

As shown in the schematic system configuration diagram of FIG. 2, the second aspect of the present invention is similar to the first aspect of the present invention shown in FIG. 1 above, except that the radiation dose rate in the main steam and water supply system is A hydrogen injection device 26 for changing the pulse shape is connected to the water supply pipe 9, and an upstream radiation detector 27, which is a radiation measuring means, is provided downstream of the hydrogen injection device 26 in the vicinity of the main steam pipe 5 or the water supply pipe 9. This upstream radiation detector
A downstream radiation detector 28 is installed further downstream than 27.

The upstream radiation detector 27 and the downstream radiation detector 28 are constructed so that their outputs are transmitted to the arithmetic processing means 29. In the arithmetic processing means 29, the upstream radiation detector 27 detects the dose rate that changes due to hydrogen injection.
Is input from the downstream radiation detector 28, the time delay between two points separated from each other is detected, and the flow velocity and flow rate are calculated.
Further, the upstream radiation detector 27 and the downstream radiation detector 28
In general, ¹⁶ N and ¹³ N, etc. are measured as the source nuclides, and the arrangement location thereof may be the purification system pipe 13.

Next, the operation of the above configuration will be described. Hydrogen is injected in a pulse shape by the hydrogen injection device 26 into the feed water which is the coolant flowing from the condenser 8 toward the reactor pressure vessel 4 through the feed water pipe 9. This hydrogen enters the reactor pressure vessel 4 through the water supply pipe 9 and reaches the reactor core 3. In the core 3, it becomes a form that easily combines with the ¹⁶ N etc. generated in the core 3 and transfers to the steam system, the ¹⁶ N etc. of this steam system increases, and the ¹⁶ N etc. in the reactor water decreases accordingly. Will be done.

In the main steam system during normal operation, nuclides such as ¹⁶ N, ¹⁵ C, ¹⁷ N and ¹³ N in reactor water which are radiation sources for the upstream radiation detector 27 and the downstream radiation detector 28 of the radiation measuring means are , Which are produced in the core 3 and are transferred to the steam system and the reactor water (purification system) depending on their chemical forms. By the way, overseas, a technique for injecting hydrogen into a reactor water supply system has been studied as a countermeasure against stress corrosion cracking of a reactor structural material, and has already been implemented in a power plant.

On the other hand, however, it is said that the amount of nuclides ¹⁶ N, ¹⁷ N, ¹³ N, etc. in the main steam system increases several times as a demerit of hydrogen injection, and ¹⁶ N produced in the core 3 is injected. This is because it is likely to combine with the generated hydrogen and transfer to the steam system, and conversely, ¹⁶ N and ¹³ N of the reactor water will decrease accordingly.

The present invention is already known (J. Magd
alinski et al: Transaction of the American Nuclear S
ociety 43 (1982)) hydrogen injection results in detail,
It was invented based on original data and experiments, and utilizes the increase and decrease of the above nuclide in each system due to hydrogen injection. Incidentally, there is data that the dose rate in the main steam system doubles when hydrogen is injected into the water supply system at a rate of only 0.5 ppm.

FIG. 6 is an explanatory view of the arrangement of radiation measuring means shown in FIG.
An example of flow rate calculation is shown in the dose characteristic curve diagram of FIG. 7. As shown in FIG. 6, an upstream radiation detector 27 is provided on the upstream side of the fluid flowing in the main steam pipe 5 along the main steam pipe 5, and the upstream radiation detector 27 is connected to the pipe distance L.
Downstream radiation detectors 28 are installed on the downstream side which is separated from each other, and the radiation data measured by each of them is transmitted to the arithmetic processing means 29 in real time. The arithmetic processing means 29 calculates the flow velocity and the flow rate from the data of the dose rate between two points separated by the pipe distance L which is changed by the hydrogen injection from the hydrogen injection device 26.

The arithmetic processing means 29 is provided with, for example, a cross-correlation meter 30, into which signals from the upstream radiation detector 27 and the downstream radiation detector 28 are input to detect downstream radiation with respect to the signals from the upstream radiation detector 27. The transit time t of the fluid is detected from the delay time of the signal of the container 28. The cross-correlation with the signal of the upstream radiation detector 27 and the signal of the downstream radiation detector 28, and the passage time t can also be analyzed by a computer.

According to this flow velocity and flow rate calculation method, when hydrogen is injected, first, the dose rate measured by the upstream radiation detector 27 on the upstream side is within a predetermined time before hydrogen injection as shown by a characteristic curve 31 in FIG. The dose rate starts to rise sharply from the averaged dose rate DB1 and becomes a constant dose rate with respect to the hydrogen injection amount. Immediately after returning to the pre-injection dose rate DB1 by stopping the injection.

On the other hand, the dose rate change by the downstream radiation detector 28 on the downstream side should be similar to the dose rate data by the upstream radiation detector 27 with a time delay corresponding to the pipe distance L of the main steam pipe 5. become. However, as shown by the characteristic curve 32, the dose rate DB2 before hydrogen injection and the maximum dose rate during injection are lower than the upstream dose rate DB1 by the time delay.

This is because the half-life of the radiation source nuclide used as the tracer is short and is attenuated while flowing down the pipe distance L. Therefore, the calculation of the arrival time between these two points is
When the dose rate F * DB1 obtained by multiplying the dose rate DB1 by an arbitrary scale factor F is reached is the upstream side arrival time t1, and the downstream side also multiplied the dose rate DB2 by the same scale factor F to reach the dose rate F * DB2. The time t2 is the downstream arrival time, the time difference (t2-t1) due to this time delay is the transit time t between two points, and the pipe distance L
The flow velocity is calculated from the following equation (1).

Flow rate = Piping distance L (cm) / Passing time t (second) (cm / second) (1) Further, the flow rate can be calculated by the equation (2). Flow rate = Velocity * Pipe cross-sectional area * Fluid density * 3600/1 x 10 ⁶ (ton / h) (2)

By periodically calculating this flow rate and evaluating it as data of secular change, it becomes possible to predict the deterioration and malfunction of the corresponding flow meter, and the difference from the existing flow meter indicated value tends to widen. In some cases, the data will be used to consider when to repair or replace. Further, the rising / falling slopes, which are changes in the dose rate in the characteristic curves 31 and 32 shown in FIG. 7, can be corrected by changing the magnification F value, and the passage time t detection accuracy can be easily improved.

The source radionuclide used for the tracer has a short half-life, so it is safe because it does not remain as radioactivity after the test is completed. Furthermore, the present invention is applicable to primary systems such as new-type converters and pressurized water reactors. Of course, it can be applied.

The following are embodiments of the claims of the first and second inventions. "The nuclear reactor monitoring device according to claim 1 or 2, wherein the source nuclide detected by the radiation measuring means is a radioactive substance generated in the nuclear reactor."

[In the arithmetic processing means for comparing the continuous measurement result of the dose rate by the radiation measuring means and the nuclear instrumentation value and the process data, the existing instrument is compared with the nuclear instrumentation value and the process data separately provided from various existing detectors. The reactor monitoring device according to claim 1, wherein the reactor monitoring device is configured to monitor the detector for signs of abnormality.

A reactor according to claim 2, characterized in that the arithmetic processing means periodically monitors and compares the measured value of the existing flow meter with the measured value so as to predict slight symptoms such as deterioration and failure of the existing flow meter. Monitoring device ".

[0046]

As described above, according to the present invention, the radiation measuring means is arranged in the vicinity of the pipes to directly and accurately measure the flow rates of the main steam system and the primary system of the nuclear power plant, and the measuring means is installed. Since the environment is good, it is easy to install and maintain the measurement means, avoid leakage of fluid and radioactivity and deterioration of the measurement means, and easily calibrate the existing measurement system to reduce the radiation exposure of workers. effective.

In the first invention, the ¹⁶ N and ¹⁵ C short-lived nuclides produced in the reactor are continuously measured from the outside of the pipe of the primary system to obtain radiation dose data during the reactor operation. Nuclear and thermal data of the plant can also be calculated. In addition, the function of existing detectors installed in the reactor and each fluid can be monitored in order to calculate the conventional plant data even when the reactor is operating.
When a detector abnormality occurs, it has an effect that it can be detected at the symptom stage.

The second aspect of the present invention has an effect that the existing flowmeter can be easily calibrated during the operation of the reactor, and abnormality such as deterioration or failure of the flowmeter can be predicted early and the time for repair or replacement can be clarified. There is.

[Brief description of drawings]

FIG. 1 is a schematic system configuration diagram of an embodiment according to the first invention.

FIG. 2 is a schematic system configuration diagram of an embodiment according to the second invention.

FIG. 3 is a block configuration diagram of a radiation measuring unit and an arithmetic processing unit according to an embodiment of the first invention.

FIG. 4 is a correlation characteristic diagram of a pipe dose rate and a main steam flow rate according to an embodiment of the first invention.

FIG. 5 is a characteristic curve diagram of a pipe dose rate, a main steam flow rate, and a heat output of an embodiment according to the first invention.

FIG. 6 is an explanatory view of the arrangement of radiation measuring means according to an embodiment of the second invention.

FIG. 7 is an explanatory characteristic curve diagram of flow rate calculation of an embodiment according to the second invention.

FIG. 8 is a schematic system configuration diagram of a conventional boiling water reactor.

[Explanation of symbols]

3 ... Reactor core, 5 ... Main steam piping, 9 ... Water supply piping, 10 ... Main steam flow meter, 11 ... In-reactor neutron detector, 13 ... Purification system piping, 14 ...
Radiation detector, 15 ... Plant data output device, 16, 29 ...
Calculation processing means, 18 ... Dose rate data memory, 22 ... Data output device, 26 ... Hydrogen injection device, 27 ... Upstream radiation detector, 28
… Downstream radiation detector, 30… Cross-correlation meter, 31… Upstream radiation detector characteristic curve, 32… Downstream radiation detector characteristic curve.

Claims (2)

[Claims] 1. A radiation detector installed in the vicinity of piping of a main steam system or a primary cooling system of a nuclear reactor, radiation measuring means for continuously measuring a radiation dose rate detected by the radiation detector, and this radiation dose. Rate and main steam flow rate, nuclear instrumentation value,
A nuclear reactor monitoring device comprising: a nuclear instrumentation value, a thermal output or a main steam flow rate of a nuclear reactor based on a correlation characteristic with a thermal output. 2. A plurality of radiation detectors installed at a predetermined distance in the vicinity of pipes of a main steam system and a primary cooling system of a nuclear reactor, and a radiation dose rate detected by the plurality of radiation detectors in time series. Radiation measuring means for continuously measuring with, a hydrogen injection device for changing the radiation dose rate to be measured in a pulsed manner, the transition time of the radiation dose rate change and the flow velocity from the predetermined separation distance of the radiation measuring means, A reactor monitoring device characterized by comprising an arithmetic processing means for calculating the flow rate of the fluid in the main steam system, the primary cooling system, etc. from the flow velocity, the pipe cross-sectional area, etc., and for performing comparative calibration with the measured value of the existing flow meter. .