JPH0415438B2 - - Google Patents

Description

[Detailed description of the invention] [Purpose of the invention] (Industrial application field) The present invention relates to a monitoring device for nuclear reactor equipment.

(Prior Art) Nuclear reactor equipment is comprised of various systems such as a high-pressure spray system and a reactor isolation cooling system, and these systems are comprised of a large number of equipment such as pumps and valves. The operating status of the equipment that makes up the reactor equipment and the process quantities of the equipment are all configured to be displayed on the control panel in the central control room. This information is displayed on the instrument or can be input by the operator.
Configured to be displayed on a CRT. Further, information that may affect safety is displayed using an alarm. Operators used this information to understand and monitor the overall condition of the equipment, maintain safety, and detect any abnormalities at an early stage.

(Problems to be Solved by the Invention) By the way, information such as the operating status of equipment and process quantities has been configured to be displayed by meters, CRTs, alarms, and the like. For this reason, for example, when operating a boiling water nuclear power generation facility, approximately 600 switches must be operated based on information from approximately 300 instruments and approximately 1,000 alarms. When monitoring each system of nuclear reactor equipment, the operating status of each device that makes up these systems was detected and displayed, and the operating status of the entire system was grasped and monitored from the operating status of these devices. . However, grasping the operating status of the entire system from the operating status of each piece of equipment requires a high level of skill and puts a heavy burden on the operator. There was a problem that made it impossible to know. For example, when an accident occurs and the operating mode is switched, it is necessary to quickly determine whether the correct flow path is formed based on the operational status of multiple valves and pumps. In addition, if the flow meter in the flow path shows zero even though the pump is operating, the flow rate can be determined by converting it from a pressure gauge, etc., but it is necessary to check whether the correct flow path is formed or not. I needed to check.

The purpose of the present invention is to organize the signals sent from the equipment side and display them in a summarized form, thereby making it easier to understand the overall status of the equipment, reducing the burden on the operator, and preventing inappropriate use. An object of the present invention is to obtain a monitoring device for nuclear reactor equipment that can reduce the possibility of erroneous operations.

[Structure of the invention]

(Means for Solving the Problems) In order to achieve the above object, in the present invention, a plurality of devices of nuclear reactor equipment, such as valves and pumps, and a first
a first detector that detects a process quantity of the system; a second detector that detects a second process quantity that is different from the first process quantity of the system; and a second detector that detects a second process quantity that is different from the first process quantity of the system; Detect whether the operating state of the system is normal or not from the valve opening/closing state signal that is the operating state signal of the device, the pump operating state signal, and the measured value of the system flow rate that is one of the process quantities, and and when the system is in normal working condition,
An estimated value of the first measured value estimated based on a predetermined calculation from a second measured value sent from a second detector that detects the second process quantity and the first process quantity. a signal processing unit that compares a first measured value sent from a first detector; and a signal processing unit that compares a first measured value sent from a first detector; displays the first measured value as the correct measured value, and if the difference between the first measured value and the estimated value of the first measured value is not within a certain range, and the operating state of the system and equipment. Provided is a monitoring device for nuclear reactor equipment characterized by comprising a display mechanism that displays the content of the determination as an abnormality message when the system is not normal.

(Function) In a device configured in this way, a detector is used that detects whether the operating state of the system is normal or not from the operating state signal of the equipment, and detects a specific process quantity and other process quantities if it is normal. Since a specific process quantity is obtained by processing the signal from the sensor, the reliability of the measured quantity can be improved.

(Example) An example of the present invention will be described below with reference to FIGS. 1 to 9.

FIG. 1 is a schematic diagram of nuclear reactor equipment according to the present invention. Boiling water nuclear power generation equipment (hereinafter referred to as equipment) 1 is equipped with a reactor pressure vessel 2 that houses a reactor core, and steam generated within this reactor pressure vessel 2 is sent to a turbine 3 to drive it. The turbine 3 drives a generator 5 to generate electricity, and the steam exiting the turbine 3 is sent to a condenser 4 where it is condensed and becomes condensed water. In addition, various types of equipment are also provided. This equipment 1 is provided with detectors for detecting process quantities or operating states of each device, and is configured so that signals from these detectors are sent to an input circuit 6. This input circuit 6 is configured to amplify each signal as necessary, convert it into a digital quantity, and send it to a signal processing circuit 7. This signal processing circuit 7 is configured to organize each signal and send these signals in a summarized form to the display mechanism 8. This display mechanism 8 has two
CRTs 9 and 10 are provided, and the first CRT9
At the top, the minimum information necessary for safe operation of the equipment 1 is summarized and displayed at all times, and the second CRT 10 is configured to display various detailed information by switching the keyboard 11. .

The first CRT 9 is configured to display information as shown in FIG. 2, for example. 12 is an item display column, and from the top, nuclear power output 1
2a, reactor water level 12b, reactor pressure 12c,
The following items are displayed: reactor water amount 12d, containment vessel pressure 12e, and radiation dose 12f. Reference numeral 13 is a measured value display column in which measured values corresponding to the above-mentioned items are displayed. This display is displayed as a bar graph 15 corresponding to the scale 14...
. . . Various limit value displays 16 . . . are displayed above as necessary. This measured value display field 13 has an upper air bar graph 15... and a digital display 17.
It is configured so that measured values are also displayed by... Reference numeral 18 is a change rate display field, which is configured to display the change rate of the measured value in correspondence with each item. This display displays an upward arrow 19 when the rate of change is positive, a downward arrow 20 when the rate of change is negative, and a horizontal line when the rate of change is zero. 21... is displayed. This rate of change display column is configured to also display the rate of change on a digital display 22.

The measured values and rate of change of each of these items are processed by the signal processing circuit 7. For example, the measured value and rate of change of the reactor water level 12b are processed as shown in FIGS. 3 and 4. displayed. As shown in FIG. 3, this reactor pressure vessel 2 is provided with narrow area water level gauges 23a, 23b, and 23c. These narrow area water level gauges 23a, 23b,
All 23c measure the water level in the reactor pressure vessel 2 in a relatively narrow range, and three of them are provided with the same structure. These narrow area water level gauges 23
Signals from a, 23b, and 23c are sent to upper and lower limit check sections 24a, 24b, and 24c, respectively.
These upper and lower limit check parts 24a, 24b, 24c
In this case, it is detected whether the signal does not exceed the measurement range of the narrow water level gauges 23a, 23b, 23c, that is, whether there is upscaling, downscaling, or an abnormality in the power supply, and if any of these abnormalities occur, Messages 25a, 25b, 25c indicating that these water level gauges 23a, 23b, 23c are out of order are displayed. If these abnormalities do not occur, the signals from the water level gauges 23a, 23b, and 23c are sent to comparators 26a, 26b, and 26c, respectively, to determine whether the absolute value of their mutual deviation is smaller than the allowable deviation value e. Determine whether If these deviations are less than the allowable deviation value e, these signals are sent to the computing units 27a, 27b,
27c, and the average values G1, G2, and G3 of the two compared signals are determined. If the deviation exceeds the allowable deviation value, the above-mentioned arithmetic units 27a, 27b,
27c, G1, G2, and G3 are each set to zero.
The signals from these computing units 27a, 27b, 27c are further sent to a computing unit 28, and these signals G
The average value of 1, G2, and G3 is determined. This average value is displayed on the aforementioned CRT 9 or the like. Signals G1, G from the arithmetic units 27a, 27b, 27c
If both 2 and G3 are zero, narrow area water level gauge 23
It is assumed that a, 23b, and 23c have all failed,
The mode shifts to the use mode of other water level measurement systems, such as wide area water level gauges installed to measure large fluctuations in the reactor water level or as back-up water level gauges for narrow area water level meters.

The water level measurement signal obtained in this way is processed as shown in Figure 4 to determine the rate of change, etc.
The above-mentioned CRT9 is displayed as a digital signal.
29 is a signal converter, and this signal converter 29
is converted into an analog signal by This signal is converted into engineering units by the engineering unit converter 30. The measurement signal converted by the engineering unit conversion section 30 is sent to the current value digital memory section 31a of the movable drawing information memory 31, and is digitally displayed on the screen of the CRT 9. Further, the signal from this engineering unit conversion section 30 is sent to the bar graph head display calculation section 32,
The starting coordinates of the bar graph display are calculated. The output from the bar graph head display section 32 is sent to the bar graph head memory section 31b and displayed on the CRT 9 as a bar graph. This signal is also sent to the current value comparison section 33, which determines whether or not the current value exceeds various limit values. If the current value exceeds the limit values, it drives the current value display color change blinking section 34. ,
The signal is displayed as the current value in the color changing/blinking memory section 31.
c and causes the display on the CRT9 to change color or blink. Next, this signal is sent to the rate of change calculation section 35.
and the rate of change over time is determined. This change rate signal is sent to the change rate memory section 31d.
Displayed on CRT9. The signal from this rate of change calculation section 35 is sent to a rate of change graphic display selection section 36, which displays an upward arrow when the rate of change is positive, a downward arrow when the rate of change is negative, and a downward arrow when the rate of change is negative. When is zero, a signal for displaying a horizontal line is output, and this signal is sent to the rate of change graphic display memory section 31e to be displayed on the CRT 9. Further, this signal is transmitted to the rate of change comparison section 37.
The system is configured to display whether the rate of change exceeds various limit values. If the limit value is exceeded, the rate of change display color change/blinking section 3
8 to the change rate display color change/blinking memory 31f and displayed on the CRT9. When this series of processing is completed, the process returns to signal conversion by the signal conversion section 29, and the measurement signals are continuously processed in this manner and displayed on the CRT 9. Note that the display of frames, characters, etc. other than the display based on the measurement signal described above is performed in advance by the frame memory section 39 of the fixed drawing information memory 39.
a and character memory 39b, thereby displaying frames, characters, etc. on the CRT 9. Furthermore, the display of each of the above data is controlled by a display request signal 40 corresponding to the operation of the keyboard 11 of the display mechanism 8.

The other CRT 10 of the display mechanism 8 is configured to display other detailed information that is not directly related to the safety of the reactor operation, such as the flow rate of the reactor isolation cooling system. Is displayed. This flow rate display also processes a plurality of measured values in the same way as described above, and shows them in a summarized form. Signal processing for displaying the flow rate of the reactor isolation cooling system is performed as shown in FIGS. 5 and 6. 41 shown in FIG. 5 is a reactor isolation cooling system. Coolant supplied from a water source such as a condensate storage tank or a pressure suppression chamber is drawn into pump 43 via inlet valve 42 . The coolant discharged from this pump 43 is supplied into the reactor pressure vessel 2 via the outlet valve 44. The pump 43 is configured to be driven by a turbine 45 driven by steam supplied from the nuclear reactor. The reactor isolation cooling system 41 configured in this manner includes a flow meter 46 for detecting the flow rate of coolant on the discharge side of the pump 43, a pressure gauge 47 for detecting the pressure on the discharge side of the pump 43, and a pressure gauge 47 for detecting the pressure on the discharge side of the pump 43. A tachometer 48 for detecting the number of rotations of the pump 43 is provided. The above pump 43
Since the characteristics of this entire system are determined in advance, the pressure on the discharge side of the pump 43 and the relationship between the rotational speed of the pump 43 and the flow rate are known in advance, and the signal from the pressure gauge 47 is
9 outputs a flow rate signal b corresponding to the flow rate, and the signal from the tachometer 48 is also outputted by a calculator 50 as a flow rate signal c corresponding to the flow rate. Note that the flow meter 46
The signal is output as is as the flow rate signal a.

These flow rate signals a, b, and c are first sent to a comparator 51 shown in FIG. ₁ , and if not, sends a signal to the signal selection unit 52, and sets the flow rate F of the reactor isolation cooling system 41 in this case to F=
Let it be a. If the deviation between the flow rate signal a and the flow rate signal b exceeds the allowable deviation value _e1 , the comparator 53
It is determined whether or not the deviation between the flow rate signal a and the flow rate signal c exceeds the allowable deviation value _e2 , and if it does not, a signal is sent to the signal selection section 52 and F=
Let it be a. If the deviation of this flow rate signal a from the flow rate signal c exceeds the allowable deviation value, a comparator 54 further determines whether the deviation between the flow rate signal b and the flow rate signal c exceeds the allowable deviation value _e3 . If it is determined that the flow rate F is not exceeded, the signal selection unit 55 sets the system flow rate F to F=
b. If the deviation between the flow rate signal b and the flow rate signal c exceeds the allowable deviation value _e3 , the signal selection unit 56 is configured to output a signal indicating that the detector has failed.

As shown in FIG. 7, the other CRT 10 of the display mechanism 8 displays an image showing a schematic configuration of a high-pressure core spray system, which is one of the emergency core cooling facilities of a nuclear reactor, for example. In Fig. 7, 2 indicates the reactor pressure vessel, 57 indicates the pressure suppression chamber, 58 indicates the condensate storage tank, 59 indicates the flow meter, 60 indicates the pressure gauge, 61 indicates the high pressure core spray pump, and 62... indicates various valves. ,
These flow meters 59, pressure gauges 60, high pressure core spray pumps 61, valves 62, etc. are given numbers to identify them. Inside the reactor pressure vessel 2, its internal pressure and reactor water level are displayed. The temperature inside the pressure suppression chamber 57 is displayed, and a pressure suppression chamber water level display section 63 is formed near the pressure suppression chamber 57.
The water level in the pressure suppression chamber 57 is displayed in a bar graph and digitally. The above condensate storage tank 58
A condensate storage tank water level display section 64 is formed near the condensate storage tank 58, and the water level in this condensate storage tank 58 is displayed in a bar graph and digitally. Above flow meter 5
A flow rate display section 65 is formed near 9, and the flow rate is displayed in a bar graph and digitally on this flow rate display section 65. The pressure is digitally displayed near the pressure gauge. An abnormality display section 66 is formed at the bottom of the screen of the CRT 10, and abnormalities occurring in the high pressure core spray system are displayed on this abnormality display section 66. An other system error display section 67 is formed in the lower right corner of this screen, and this other system error display section 67 displays information when an error occurs in a system other than the system currently displayed on this screen. It is configured to display this anomaly.

This display is made as shown in FIG. Reference numeral 68 indicates a high-pressure reactor core spray system, and 2 is the aforementioned reactor pressure vessel. A condensate storage tank 58 serves as a water source for the high pressure core spray system 68. 57 is a pressure suppression chamber, and similarly this high pressure core spray system 6
This is the water source for No. 8. Water level gauges 69 and 70 are provided in the condensate storage tank 58 and the pressure suppression chamber 57, respectively. 61 is a high-pressure core spray pump, and the suction side of this high-pressure core spray pump 61 is connected to the condensate storage tank 58 and the pressure suppression chamber 5 through suction valves 71 and 72, respectively.
7 is connected. The discharge side of this high-pressure core spray pump 61 is connected to a high-pressure spray nozzle (not shown) in the reactor pressure vessel 2 via an injection valve 73 and a check valve 74, in case a loss of coolant accident occurs. In some cases, the reactor is configured to inject coolant into the core. The above high pressure core spray pump 6
1, a pressure gauge 60 and a flow meter 59 are provided on the discharge side. This high pressure core spray pump 61
A minimum flow line 75 for preventing cavitation is branched and connected to the discharge side of the pump, and a minimum flow valve 76 is provided in the middle of the minimum flow line 75. Test piping 77 and 78 for testing and maintenance are provided branching from the discharge side of the high pressure furnace spray pump 61,
Test valves 79 and 80 are provided in the middle of these test pipes 77 and 78, respectively. These test valves 79 and 80 are closed during normal operation. Signals from the water level gauges 69 and 70 in the condensate storage tank 58 and pressure suppression chamber 57 and signals corresponding to the open and closed states of the suction valves 71 and 72 are input to a water source check section 81, respectively. The water source check section 81 determines whether the water levels in the condensate storage tank 58 and the pressure suppression chamber 57 are above a predetermined water level, and whether the suction valves 71 and 72 are open. When the water level is below the predetermined water level or when the suction valves 71 and 72 are not opened, a message 82 indicating that an abnormality has occurred in the water source is sent.
It is displayed on the abnormality display section 66 on the screen of the CRT 10.
If the above-mentioned abnormality does not occur, a signal indicating that it is normal is sent to the comparison calculation section 83. A signal corresponding to the operating state of the high pressure core spray pump 61 is sent to a pump operating state check section 84. This pump operation status check section 84 determines whether the high pressure core spray pump 61 is operating normally or not. If it is not operating normally, a message 85 is sent to the effect that the high pressure core spray pump 61 is abnormal.
is displayed on the abnormality display section 66 on the screen of the CRT 10, and if the CRT 10 is operating normally, a signal to that effect is sent to the comparison calculation section 83. Signals corresponding to the open/close states of the injection valve 73, check valve 74, test valves 79, 80, etc. on the discharge side of the high pressure core spray pump 61 are sent to the valve line up check section 86, and these valves are It is determined whether or not they are arranged in a predetermined open/closed state, that is, whether they are in a predetermined line-up. If these valves are not aligned in the specified opening/closing state, a message 87 will be sent to the CRT indicating that there is an abnormality in the valve line-up.
10 on the abnormality display section 66, and if the valve line-up is normal, a signal to that effect is sent to the comparison calculation section 83. This comparison calculation section 83
Then, it is determined whether all these signals are normal or not, and if all are not normal, the calculation unit 88 displays on the flow rate display unit 65 of the CRT 10 that the flow rate of this high pressure core spray system 68 is zero. , if everything is normal, a signal to that effect is sent to the minimum flow valve check section 89. This minimum flow valve check section 89 includes the minimum flow valve 7.
A signal corresponding to the open/closed state of the valve 6 is sent, and it is determined whether the minimum flow valve 76 is normal or closed. If this minimum flow valve 76 is normally closed, the calculation unit 9
The flow rate signal measured by the flow meter 59 by
_S1 is the first flow rate signal _G1 , and a converted flow rate signal is obtained by converting the pressure measured by the pressure gauge 60 into a flow rate.
Let F _QH be the second flow rate signal _G2 . When the minimum flow valve 76 is open, the calculation unit 91 subtracts the flow rate K flowing through the minimum flow pipe 75 from the flow rate signal S ₁ as the first flow rate signal G ₁ and the converted flow rate signal F _QH . The value obtained by subtracting the flow rate K from the above is defined as the second flow rate signal _G2 . This first
It is determined whether the deviation between the flow rate signal G ₁ and the second flow rate signal G ₂ exceeds the allowable deviation value e, and if it does not, the calculation unit 93 CRT10 indicates that the flow rate is G ₁ .
If the flow rate exceeds G 1 or G 2 , the arithmetic unit 94 displays on the flow rate display unit 65 of the CRT 10 that the flow rate is the smaller of G ₁ or G ₂ .

The above-mentioned process quantity measurement signals for each system and detection signals for the operating status of each system are comprehensively processed in the entire nuclear reactor equipment, and are used indirectly to measure the process quantities and other things for the entire equipment. For example, FIG. 9 shows a case where the reactor water level is indirectly measured from the measurement results of the process quantities of each system. That is, 2 is the above-mentioned nuclear reactor pressure vessel, and the reactor core 95 is housed inside this vessel. The coolant in this reactor pressure vessel 2 is supplied to jet pumps 96... and recirculation pumps 97.
... is circulated through the core 95. The steam generated within the reactor pressure vessel 2 is configured to pass through a main steam pipe 98 and be sent to the turbine 3 via a turbine throttle valve 99. Water is supplied into the reactor pressure vessel 2 through a water supply pipe 100. 101 is a control rod drive mechanism for driving control rods (not shown), and is driven by water pressure. The water level in the reactor pressure vessel 2 is configured to be detected by a reactor water level gauge 102. The above reactor pressure vessel 2
By combining all the coolant flow rates related to the coolant flowing into the reactor pressure vessel 2, and by combining the coolant flow rates of all systems related to the coolant flowing out from this reactor pressure vessel 2, Calculate the water level in the reactor pressure vessel 2 from the amount of coolant flowing into the reactor pressure vessel 2, the amount of coolant flowing out, and the amount of coolant retained in the reactor pressure vessel 2 in the initial state, It is used as a backup water level measurement system for the reactor water level gauge 102. That is, the flow rate Q ₁ of the coolant flowing out from the main steam pipe 98 is measured based on the state of the turbine system such as the opening degree of the turbine throttle valve 99 and the operating state of the turbine 3 , and is sent to the signal processing circuit 7 . Further, the flow rate Q ₂ of the coolant flowing in from the water supply pipe 100 is measured by the flow meter 103 . In addition, in the reactor isolation cooling system 41, the flow rate Q ₃ of the coolant extracted from the main steam pipe 98 to drive the turbine 45
and the flow rate of coolant injected by pump 43
Find Q ₄ . In addition, the flow rate Q ₅ of the coolant flowing out of the reactor water purification system 104 and the flow rate Q ₆ of the coolant flowing into the reactor water purification system 104 are determined. In addition, the control rod drive mechanism 101
Find the flow rate _Q7 of the coolant flowing in from... Also,
The flow rate Q ₈ of the coolant injected from the boric acid water injection system 105 is determined. Furthermore, the flow rate Q ₉ flowing out and the flow rate Q ₁₀ flowing into the residual heat removal system 106 are determined. Furthermore, the flow rates Q ₁₁ , Q ₁₂ , and Q ₁₃ of the coolant flowing from the high pressure core spray system 68, the low pressure core spray system 107, and the low pressure injection system 108 are determined. The signals corresponding to these flow rates Q ₁ , Q ₂ , Q _{3 ,} . In this case, some of the coolant flowing out is in the vapor phase, but all of these are converted to the flow rate of the coolant in the liquid phase. When measuring the flow rate of each of the above systems, if there are multiple overlapping detectors, compare the signals from each detector as described above, exclude those exceeding the allowable deviation value, and calculate the remaining signals. Calculate the average value, and if multiple signals that can be converted to flow rate are obtained, convert and compare the signals as described above, exclude those exceeding the allowable deviation value, and calculate the average value of the remaining signals. and improve the accuracy of flow measurement. Furthermore, the enthalpy of the inflowing and outflowing coolant can be determined from the pressure and temperature, and this enthalpy can be multiplied by the flow rate to calculate the energy flowing into and out of the reactor pressure vessel 2, and the output of the reactor can also be determined. .

One embodiment of the present invention configured as described above detects the operating status of each system of nuclear power generation equipment,
Compare and synthesize the operating status of each device that makes up the system,
The operating status of the entire system, that is, whether or not an abnormality has occurred in the system is displayed. Therefore, the operator can immediately grasp the status of the entire system just by monitoring the display of the operating status of the entire system, reducing the burden on the operator and ensuring that abnormalities in the entire system are not overlooked. .

Also, when determining a specific process quantity (for example, flow rate) from a detector that measures another process quantity (for example, pressure), the pressure measured by the pressure gauge is converted to flow rate after inspecting the operating status of the system. , the risk of calculating an incorrect flow rate can be reduced.

Furthermore, since we compare and diagnose detectors with different measurement principles (for example, a flow meter and a pressure gauge), even if a principle defect occurs in the flow meter itself, the flow rate can be measured from the operating status of the pressure gauge and the system. can improve the health of your monitoring.

〔Effect of the invention〕

According to the present invention, when measuring a specific process quantity, the specific process is determined based on a signal from a detector that detects the specific process quantity, a signal from a detector that detects other process quantities, and the operating state of the system. Since the quantity is determined, the reliability of the measured quantity can be improved.

In addition, since signals from detectors with different measurement principles are processed, even if a principle defect occurs in the detector that detects a specific process quantity, the signals from the detectors that detect other process quantities A specific process quantity can be calculated from the system operating state.

Therefore, according to the present invention, a large amount of signals sent from the equipment side are organized and displayed in a summarized form, and this summarized information is highly reliable. Operators can easily grasp the status of the entire equipment.

[Brief explanation of drawings]

The drawings show one embodiment of the present invention; FIG. 1 is an overall schematic diagram, FIG. 2 is a diagram showing an example of the display of the display mechanism, FIG. 3 is a flowchart of signal processing in the case of reactor water level measurement, and FIG. Figure 4 is a flowchart of signal processing for the same display.
Figure 5 is a configuration diagram of the reactor isolation cooling system, Figure 6 is a flowchart for processing the flow rate signal of the reactor isolation cooling system, Figure 7 is a diagram showing another display example of the display mechanism, FIG. 8 is a flowchart of signal processing when detecting the state of the high-pressure core spray system, and FIG. 9 is a system diagram of the reactor illustrating the case where the reactor water level is indirectly measured. 1... Nuclear reactor equipment, 2... Reactor pressure vessel, 3
... Turbine, 6 ... Signal input circuit, 7 ... Signal processing circuit, 8 ... Display mechanism, 9, 10 ... CRT,
23a, 23b, 23c...Narrow area water level gauge, 41...
...Reactor isolation cooling system, 46...Flow rate system, 47...
...Pressure gauge, 48... Rotation meter, 57... Pressure suppression chamber, 58... Condensate storage tank, 59... Flow meter,
60...Pressure gauge, 69,70...Water level gauge, 71,
72...Suction valve, 73...Injection valve, 76...Minimum flow valve.

Claims (1)

[Scope of Claims] 1. Valves and pumps that are a plurality of devices of nuclear reactor equipment; a first detector that detects a first process amount of a system configured by these devices; a second detector that detects a second process quantity different from the first process quantity; and a valve opening/closing state signal and a pump operating state signal that are operating state signals of the equipment sent from the equipment. Detecting whether or not the operating state of the system is normal from the measured value of the system flow rate, which is one of the process quantities, and when the operating state of the equipment and the system is normal, the second process quantity is determined. a first detector that detects an estimated value of the first measured value estimated based on a predetermined calculation from a second measured value sent from a second detector that detects the first process quantity; a signal processing unit that compares the first measured value sent from the first measured value; and when the difference between the first measured value and the estimated value of the first measured value is within a certain range, the first measured value is If the value is displayed as a correct measurement value, and the difference between the first measurement value and the estimated value of the first measurement value is not within a certain range, and if the operating state of the system or equipment is not normal, A monitoring device for nuclear reactor equipment, comprising a display mechanism that displays the determination content as an abnormality message. 2. The nuclear reactor according to claim 1, wherein the display mechanism displays the operating status of the entire system and also displays the operating status of each of the devices constituting the system. Equipment monitoring device. 3. A patent claim characterized in that, when an abnormality occurs in a system other than the currently displayed system, the display mechanism displays on a part of the screen that an abnormality has occurred in the other system. A monitoring device for nuclear reactor equipment according to item 1.