JPH02285290A - Reactor state visualization system - Google Patents

Abstract

PURPOSE:To reduce a burden of operators by providing the system with an animation operation part for animating an analysis result from a reactor characteristic analysis part and an animation display for displaying information from the animation operation part. CONSTITUTION:A reactor state visualization system consists of a reactor characteristic analysis part 13, a neutron dynamic characteristic analysis part 15, a memory 17, an animation display 21 and the like. The analysis part 13 is allowed to analyze a dynamic characteristic on thermohydraulic behavior of a reactor pressure vessel on the basis of a measured value signal or an operation signal and evaluate a reactor state in a real time. The analysis part 15 is allowed to substitute a dynamic characteristic value on the thermohydraulic behavior analyzed in the analysis part 13 for a neutron dynamic characteristic equation to find reactivity so as to calculate the level of a neutron flux in a real time. The memory 17 is allowed to store a transient change of the thermohydraulic behavior dynamic characteristic analyzed in the analysis part 13 in a time series. The animation operation part 19 is allowed to animate the thermohydraulic behavior stored in the memory 17. Its result is displayed in the display 21.

Description

[Detailed Description of the Invention] [Object of the Invention] (Industrial Application Field) This invention relates to the reactor condition of a nuclear power plant, especially y
The present invention relates to a nuclear reactor status visualization system that suitably displays the status of the nuclear reactor at the time of delivery.

(Prior Art) A system for monitoring the plant status of a nuclear power plant is comprised of a measuring section 1, a performance section 3, and a display section 5, as shown in FIG. The measurement unit 1 measures the reactor pressure, the differential pressure at the core inlet, the feed water temperature, the feed water flow m, the neutron flux level, and the like.

The calculation unit 3 calculates the core coolant flow rate, reactor output, etc. based on these measured values. Various plant states calculated by the calculation unit 3 or measured by the measurement unit 1 are displayed on the display unit 5 in real time.

The display screen of the display unit 5 is, for example, as shown in FIG. In FIG. 5, the reactor pressure, reactor coolant flow ffi#, reactor output, and generator output are displayed as numerical values and trends. By displaying the plant status in this manner, operators can deepen their understanding of the transition of arbitrary phenomena within the reactor pressure vessel. In addition, the fifth
The trend display in the figure shows the trend at the time of IS modification when cold members were lost due to a rupture in the recirculation system piping.

(Problem to be Solved by the Invention) However, although the display unit 5 displays plant status such as reactor pressure and core coolant temperature numerically or graphically, No information regarding mechanical behavior is displayed. Information regarding this thermo-hydraulic behavior includes, for example, changes in the void ratio of the working fluid in various parts within the reactor pressure vessel, the flow state of the working fluid, and the surface angle of the fuel rod cladding tube.

Therefore, regarding these thermo-hydraulic behaviors, operators must individually estimate them from the reactor pressure, core coolant flow rate, etc. displayed on the display unit 5. For example, when the reactor pressure is high, the void ratio of the working fluid is low. Therefore, in such conventional reactor condition monitoring systems, the level of understanding of the transition of phenomena related to thermo-hydraulic behavior inside the reactor pressure vessel varies depending on the operator, and sometimes that understanding is inaccurate. In the worst case scenario, there is a risk that the operator will make an erroneous operation.

This invention was made in consideration of the above facts,
A reactor status visualization system that allows each operator to clearly understand the transition of phenomena related to thermo-hydraulic behavior within the reactor pressure vessel, reduces the burden on operators, and prevents operator errors. The purpose is to provide

[Structure of the invention]

(Means for Solving the Problems) This invention analyzes the dynamic characteristics of the thermo-hydraulic behavior in the reactor pressure vessel by inputting data on the thermo-hydraulic behavior in the reactor pressure vessel, and A reactor characteristics analysis section that evaluates the reactor status in real time, an animation calculation section that animates the analysis conclusion from this reactor characteristics analysis section, and an animation display that displays information from this illusion performance section. It has a section.

(Operation) Therefore, the nuclear reactor state visualization system according to the present invention allows operators to directly visually recognize the transition of phenomena related to thermal hydraulic behavior within the reactor pressure vessel.

(Example) Embodiments of the present invention will be described below based on the drawings.

FIG. 1 is a block diagram showing an embodiment of the nuclear reactor state visualization system according to the present invention.

The reactor status visualization system 11 is a reactor characteristics analysis section 13
, a neutron dynamic characteristic analysis section 15, a storage section 17, an animation calculation section 19, and an animation display section 21.

The reactor characteristics analysis section 13 includes a reactor measurement system 23 and a control tI! A panel 25 is electrically connected, and changes in reactor main steam pipe pressure, feed water temperature changes, feed water temperature changes, etc. measured by the reactor measurement system 23 are input as measurement value signals. Further, from the control wJ125, changes in the plant state due to operation H operations (reactor recirculation pump trip, υ control rod insertion, etc.) are input as operation signals. These measured value signals and operation signals are data regarding the thermo-hydraulic behavior within the reactor pressure vessel.

The reactor characteristic analysis unit 13 analyzes the dynamic characteristics regarding the thermal-hydraulic behavior within the reactor pressure vessel based on the measurement value signal and the operation signal, and evaluates the state of the reactor 2 in real time.

The dynamic characteristics related to thermo-hydraulic behavior include the void ratio of the working fluid in the reactor pressure vessel, the flow state of the working fluid (core coolant flow rate and core coolant temperature), fuel rod cladding surface degree, and atomic This is the furnace pressure distribution.

The neutron dynamic characteristic analysis unit 15 is electrically connected to the reactor characteristic analysis unit 13 and is also connected to the control panel a 25.The neutron dynamic characteristic analysis unit 15 is connected to the reactor characteristic analysis unit 13 The reactivity is obtained by substituting the value of the dynamic characteristic regarding the thermal-hydraulic behavior in the reactor pressure vessel analyzed in the neutron dynamic characteristic equation, and this reactivity and the operation signal from the control panel 25 are taken into consideration. ,・Calculate the neutron flux level in real time.Furthermore, calculate the reactor power distribution and reactor output in real time from the calculated neutron flux level.

The reactor power and the like calculated by the reactor dynamic characteristics analysis section 15 are input to the reactor characteristics analysis section 13.

In addition to this, the nuclear reactor characteristics analysis unit 13 also has minute time (for example, 1/100 second) I! After passing, the measurement value signal from the reactor measurement system 23 and the operation signal from the control panel 25 of the same minute time lapse are transmitted to the reactor measurement meter 23.1jllFD system 25, respectively.
Input from The reactor characteristics analysis unit 13 analyzes the inside of the reactor pressure vessel after a short period of time has elapsed since the first analysis of the dynamic characteristics related to the thermo-hydraulic behavior, based on the measured value signals and operation signals such as the reactor output. Calculate the dynamic characteristics regarding the thermo-hydraulic behavior of Here, the minute time is the time required for processing by the neutron dynamic characteristic analysis section 15. The reactor characteristic analysis unit 13 analyzes transient changes in dynamic characteristics related to thermo-hydraulic behavior while advancing time in this manner.

The storage unit 17 stores in time series the transient changes in dynamic characteristics related to the thermal-hydraulic behavior within the reactor pressure vessel analyzed by the reactor characteristic analysis unit 13.

In addition, the animation calculation unit 11 stores information on thermo-hydraulics in the reactor pressure vessel stored in the storage unit 17 (the flow state of the working fluid in the reactor pressure vessel, the void ratio of the working fluid, reactor pressure, fuel rod cladding surface temperature)
The result is displayed on the animation display section 21.

In other words, as shown in FIG. are displayed as simulated fluid particles 29. The flow state of the working fluid within the reactor pressure vessel 27 is displayed by the movement of the simulated fluid particles 29.

The change in the void ratio of the working fluid in the reactor pressure vessel 27 displayed on the animation display section 21 is represented by a change in the color of two of the large number of simulated fluid particles displayed on the animation display section 21. That is, when α>0.99 and the working fluid is steam, the simulated fluid particles 29 are displayed in red. In addition, when the working fluid is a gas-liquid two-phase flow, 0.8〈α≦
When the value is 0.99, the simulated fluid particles 29 are pink, and when the value is 0.0
When 5<α≦0.8, the simulated fluid particles 29 are displayed in light blue. Roughly speaking, when α≦0.05 and the working fluid is water, the simulated fluid particles 29 are displayed in blue.

For example, when the low-pressure core spray system (LP01) or the low-pressure injection system (LPCI) is activated, the color of the simulated fluid particles 29 near their nozzles changes to a color with a low void ratio.

In this FIG. 2(A), the reference numeral 30 indicates the reactor core, of which the reference numeral 30A indicates the high-power fuel assembly, and the reference numeral 30B indicates the reactor core.
denote the average power fuel assembly, respectively. Also, code 31
33 is a jet pump, and 35 is a lower tie plate.

Changes in the reactor pressure inside the reactor pressure vessel displayed on the animation display section 21 are as shown in FIG. 2(B).
As time passes after the occurrence of the abnormal event, numerical values are displayed on the animation display section 21 in real time. Alternatively, the change in reactor pressure may be displayed as a graph of the time change after the occurrence of the abnormal event.

The surface temperature of the fuel rod cladding tube displayed on the animation display section 21 is represented by the surface temperature of the cladding tube of the high-output fuel assembly 30A, as shown in FIG. 2(C). This second
In Figure (C), the horizontal axis represents the temperature (K), and the vertical axis represents the height H of the high-output fuel assembly 30A, and this height H is divided at regular intervals. Then, for each divided height of the fuel assembly, the bar graph is displayed so that if the temperature is high, the bar graph moves to the right in the figure, and if the temperature is low, it moves to the left. In this FIG. 2(C), TSAT indicates the saturation temperature of water, which is the working fluid.

Next, the effect will be explained.

When the nuclear reactor is operating normally, the reactor characteristics analysis unit 13 analyzes the dynamic characteristics regarding the thermo-hydraulic behavior in the steady state of the reactor at least once, and stores the results in the storage unit 17. .

In this case, the dynamic characteristics regarding the thermo-hydraulic behavior are v
This is the initial value when calculating the J characteristic. Therefore, the dynamic characteristics related to the thermo-hydraulic behavior in the steady state do not necessarily need to be displayed on the animation display section 21, but are displayed on the animation display section 21 if requested by the operator.

FIG. 3(A) shows the color change of the simulated fluid particles 29 indicating the void ratio of the working fluid in the reactor pressure vessel in a steady state. The simulated fluid particle 29 located at the top of the reactor pressure chamber 127 displayed on the animation display section 21 is displayed in red, indicating that the void ratio of the working fluid is the highest. In addition, the position of the reactor core 30 and the simulated fluid particles 29 located above it are displayed in pink,
Furthermore, the dummy particles 29 located in most of the reactor core 30 and in the upper part of the reactor pressure vessel 27 are displayed in light blue. Furthermore, the simulated fluid particles 29 located at the lower part of the reactor pressure vessel 27 and the outer periphery of the shroud 31 are displayed in blue, indicating that the void ratio of the working fluid is the lowest.

Next, a case where an abnormal transient event occurs or an accident such as a pipe breakage occurs due to some abnormality will be described.

In this case, the measured value measured by the reactor measurement system 23 changes significantly from the steady state, and furthermore, the plant operating state also changes significantly due to the plant operating operation performed by the operator. Therefore, the reactor characteristic analysis unit 13 uses the dynamic characteristics related to the steady-state thermo-hydraulic behavior as the first W1vi stored in the storage unit 17, and roughly calculates the measurement value signal from the reactor measurement system 23 and the II ml Dynamic characteristics related to thermo-hydraulic behavior are analyzed based on the operation signal from the panel 25. The neutron dynamic characteristic analysis unit 15 performs real-time processing based on the analysis results to calculate reactor pressure and the like. The reactor characteristics analysis unit 13 uses the reactor pressure calculated by the neutron dynamic characteristics analysis unit 15, the measured value signal from the reactor measurement system 23 after a short time has elapsed, and the operation signal from the control m panel 25. , minute rII
Dynamic characteristics regarding thermo-hydraulic behavior after the lapse of time are determined. In this manner, the reactor characteristic analysis unit 13 analyzes transient changes in dynamic characteristics regarding thermo-hydraulic behavior.

The analysis results of the dynamic characteristics regarding the thermo-hydraulic behavior during this abnormality are sequentially output from the reactor characteristics analysis unit 13 to the animation calculation unit 19 via the storage unit 17 and are animated, and the animation display f! It is displayed on 1s21. Among these, Figure 3 (B) shows the void ratio of the DR body in the reactor pressure vessel after the rupture accident of the reactor-subcooling system piping until just before the low pressure injection system (LPCI) is activated. , which is displayed as a color change of the simulated fluid particles 29.

The imitation particles 29 located above and in the vicinity of the core 30 displayed on the animation display section 21 are displayed in pink. Furthermore, the simulated fluid particles 29 located below the lower tie plate 35 are also displayed in pink. Furthermore, the simulated fluid particles 29 located at the lower part of the reactor pressure vessel 27 and near the shroud 31 are displayed in light blue. In addition, the simulated fluid particles 29 located in the upper part of the reactor pressure vessel 27 and the outer periphery of the shroud 31 are displayed in red, indicating that these parts have the highest void ratio.

Figure 3 (C) shows 1. The change in void ratio from the activation of the low pressure injection system (LPCI) until the core is re-submerged is displayed as a color change of the simulated fluid particles 29.

The simulated fluid particles 29 located in the reactor core 30, the lower part of the reactor pressure vessel 27, and the jet pump 33, which are displayed on the animation display section 21, are displayed in light blue. Furthermore, the simulated fluid particles 29 located below the lower tie plate 35, on the lower outer periphery of the jet pump 33, and above the jet pump 33 on the high-output fuel assembly 30A side are displayed in pink. Further, the simulated fluid particles 29 located near the bottom of the shroud 31 are displayed in blue, indicating that the void ratio is low. Further, the upper part of the reactor pressure vessel 27 and the outer circumference of the shroud 31 are displayed in red, indicating that the void ratio is the highest.

According to the above embodiment, the reactor characteristics analysis section 13 analyzes the dynamic characteristics related to the thermo-hydraulic behavior within the reactor pressure vessel, and the animation calculation section 19 animates the analysis results. Since it is displayed on the animation display section 21, the operator can visually recognize the transition of the phenomenon related to the thermo-hydraulic behavior inside the reactor pressure vessel. As a result, operators can clearly understand the thermal-hydraulic behavior within the reactor pressure vessel. Therefore, compared to conventional plant status monitoring systems that have to estimate the reactor status from reactor pressure, reactor coolant flow m, etc., the burden on operators can be significantly reduced and operator errors can be prevented. You can also.

〔Effect of the invention〕

As described above, according to the reactor state visualization system according to the present invention, the reactor characteristic analysis section inputs data regarding the thermal hydraulic behavior within the reactor pressure vessel and The dynamic characteristics related to thermo-hydraulic behavior are analyzed and the reactor status is evaluated in real time.The animation calculation section animates the analysis results from the reactor characteristics analysis section, and the animation display section performs animation calculations. By displaying information from the reactor pressure vessel, operators can visually clearly understand the transition of phenomena related to thermal-hydraulic behavior within the reactor pressure vessel, especially in transient conditions. As a result, it is possible to contribute to reducing the burden on the operator and preventing erroneous operations.

[Brief explanation of drawings]

FIG. 1 is a block diagram showing an embodiment of the nuclear reactor status visualization system according to the present invention, and FIGS. 2(A), (B), (C)
) is a diagram showing the screen of the animation display section, and Figure 3 (A
), (8), and (C) are animation display screens showing changes in the void ratio of the working fluid in the reactor pressure vessel before and after the reactor-subcooled W system piping ruptures as color changes in simulated fluid particles. 4 is a block diagram showing a conventional nuclear reactor status visualization system, and FIG. 5 is a display screen of a display unit in the conventional nuclear reactor status visualization system. 11... Reactor status visualization system, 13... Reactor characteristics analysis section, 15... Neutron dynamic characteristics analysis section, 19.
... Animation calculation section, 21... Animation display section, 27... Reactor pressure vessel displayed on the animation display section, 29... Simulated fluid particles displayed on the animation display section, α... Void rate. Figure 1 Applicant's agent Hisashi Hatano 0M T-0,
0 (5) Clear - 5T Haze 50aO (S) (A) (B) Figure time T-58,84 (S) (C) Figure Figure

Claims (1)

[Claims] 1. Data regarding the thermo-hydraulic behavior within the reactor pressure vessel is input and the dynamic characteristics regarding the thermo-hydraulic behavior within the reactor pressure vessel are analyzed to determine the reactor state in real time. It has a reactor characteristic analysis section for evaluation, an animation calculation section for animating the analysis results from the reactor characteristic analysis section, and an animation display section for displaying information from the animation calculation section. A nuclear reactor status visualization system featuring: 2. The dynamic characteristics related to the thermo-hydraulic behavior to be animated are the void ratio of the working fluid in the reactor pressure vessel, the flow state of the working fluid, and the surface temperature of the fuel rod cladding tube. Reactor status visualization system. 3. The void ratio of the working fluid and its flow state are displayed by simulating the fluid particles of the working fluid in the reactor pressure vessel on an animation display section, and by changing the color of the simulated fluid particles and their movement. Item 2. Nuclear reactor status visualization system according to item 2.