JPS60100795A - Method and device for predicting state of inside of reactor on accident of nuclear reactor - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

[Detailed Description of the Invention] [Field of Application of the Invention] The present invention relates to a method and a device for predicting changes in the state inside a boiling water reactor during an accident, and in particular to a method and apparatus for predicting changes in the state inside the reactor during an accident in a boiling water reactor. The present invention relates to a method and apparatus for estimating and predicting changes that will occur in a furnace in the future.

[Background of the invention]

At nuclear power plants, in order to avoid accidents, species prevention and
Operational guidance is now provided to safely shut down the furnace. In this way, nuclear power plants have double and triple safety measures in place, but in the wake of the Three Mile Island accident, it is necessary to develop equipment that can predict changes in the reactor due to an accident should an accident occur. It is said to be necessary. The purpose of this is to estimate the approximate scale of an accident such as a pipe rupture and determine the most appropriate operating procedure. It is necessary to predict the dynamic behavior of

Various dynamic simulators have been developed to simulate the dynamic behavior of conditions inside a nuclear reactor. However, many of these dynamic simulators require calculation time that is longer than the speed at which the phenomenon progresses, and even some simulators used for operator training are barely able to calculate at the same speed as the actual progress of the phenomenon. , that is, it can only perform real-time calculations. Therefore, it is impossible for this type of simulator to predict the development of phenomena caused by an accident in advance when an accident occurs.

In order to improve this problem, efforts have been made to simplify the model for simulating nuclear reactor conditions and to develop simulators that are several times faster than real time. Although this method provides high-speed computation, it comes at the cost of computation accuracy during prediction due to simplification. Predicting the evolution of a nuclear reactor accident requires prediction over a long period of two to three hours, and it can be said that a prediction method that does not guarantee sufficient accuracy will not function as a prediction. On the other hand, as a method for improving prediction accuracy, there is a method of adjusting some parameters of a prediction simulator based on actual data, but this method is also not useful for the above-mentioned long-term prediction.

[Purpose of the invention]

The purpose of the present invention is to provide a system that can predict the development of phenomena for two to three hours in about 10 to 20 minutes while maintaining sufficient accuracy when an accident occurs in a boiling water reactor. An object of the present invention is to provide a method and device for predicting the state inside a reactor.

[Summary of the invention]

The present invention uses a necessary and sufficient simplified model based on measured data of the nuclear reactor plant at the time of the accident to estimate the scale of the accident in a short time, and also estimates the scale of the accident with accuracy based on the estimated parameters. Initial settings are made with a model that has been simplified to the extent that it can be fully guaranteed, and based on this, it is possible to predict the long-term development of phenomena caused by accidents.

[Embodiments of the invention]

Hereinafter, the present invention will be explained based on FIGS. 1 to 8.

First, FIG. 1 is a diagram illustrating a system configuration of an example of an in-reactor state prediction device for a boiling water nuclear reactor according to the present invention.

Process quantities from the nuclear reactor plant 1 are taken into the input device 2 of this system. These process quantities include the reactor water level, reactor pressure, reactor output, recirculation flow rate control signal, control rod drive signal, and other signals used in the accident scale estimating device.
and core flow rate, main steam flow rate, feed water flow rate, containment vessel temperature,
It consists of signals 9 such as the collar recirculation flow rate for the storage unit and the flow rate signals of various emergency cooling systems. The accident scale estimating device 3 uses the process signal 8 to estimate, for example, the scale of a pipe rupture. As described later, the simulation model used with this device is highly simplified in order to estimate the scale of an accident in a short time, and the items for estimating the scale of a pipe rupture must be limited to the rupture area and rupture quality. I don't get it. Here, the rupture area is the cross-sectional area where the contents leak from the rupture part, and the rupture quality represents the content of water in the leaked material, and is a numerical value between 0 and 1. In other words, it is 0 if there is no water, and 1 if there is only water. On the other hand, in order to predict accidents and subsequent long-term phenomena, the prediction device 5 is equipped with a relatively detailed simulation model that can simulate all of the process quantities described above. This model also simulates various types of piping. Therefore, in order to start prediction with this device, it is necessary to switch from the already obtained fracture area and fracture quality to a fracture area that matches the actual fracture position, and for this purpose, the fracture position estimation device 4 and the database 7 is available.

The predicted values of the state inside the nuclear reactor or the plant obtained by the prediction device 5 are displayed in chronological order on a display device 6 such as a CRT to support the judgment of the operator.

Next, each device of the reactor state prediction device according to this system will be explained. FIG. 2 shows the configuration of the fracture size estimating device 3. The scale estimation process signal 8 is input to the state determining section 10. Based on these process signals 8, the status determination unit 10 determines whether changes in the reactor water level, reactor pressure, etc. are due to an accident. As an example of the determination, for example, a change in the furnace output is defined as an event person, and is expressed as A=(there is a change in the furnace output). Similarly, B = (with a change in the reactor water level), C = (with a change in the reactor pressure), D = (with a change in the recirculation flow rate control signal), and E = (with a change in the control rod drive signal).

When AAB△(C△(DAE)) is true, that is, AAB△(C・△(DAE))=1 (1), the determination unit 10 is open, and the process signal 8 is input to the storage device 11. be done. In the above pool algebraic expression, △ indicates AND, and . indicates NOT.

The storage device 11 stores time-series data of the process signal 8. Further, the process signal 8 is outputted to the current state estimating section 12 and the comparing section 15 in response to a trigger signal 14 applied to this system by an operator. In the determination unit 10, even if the above logical determination is negative, the circuit is opened approximately once every hour. In other words, even if no accident occurs, the determination unit 10 is not opened.
Process signal 8 is input to storage device 11 . In this case, the determination section 10 closes in about 2 to 3 minutes, and the signals when the logical determination is negative are averaged and stored as one value. This value is similarly input to the current state estimating section 12 as a steady state signal 16 by a trigger signal.The current state estimating section 12 automatically sets the initial value of the model from the value of the steady state signal 16.

At the same time, among the process signals 8, the cooling system flow rate, enthalpy, and recirculation flow rate are input, and the fracture quality and fracture area are set to assumed values, and the estimation unit 12 uses the model to calculate the reactor water level, Calculate furnace pressure. This calculation is performed for several minutes, and the comparing section 15 compares the actually measured reactor water level and furnace pressure corresponding to the simulation period with the values obtained by the above calculation.

The actual measured values of the reactor water level and reactor pressure are now XI(t), respectively.
, Xz(L), the calculated values of reactor water level and reactor pressure are expressed as yt(t
).

If yz(t), ...(2) 1 Simulation I/Function to sequentially change the assumed values of the fracture area and fracture quality so as to give the fracture area and fracture quality that minimize the evaluation quantity. Optimal estimator i
α and β of the above-mentioned evaluation quantities are non-negative appropriate values, tO is the time when the calculation is started by the trigger signal 14, T is the time to evaluate the actual measured value and the calculated value, As mentioned earlier, it only takes about a few minutes.

The optimal estimator 13 can use a nonlinear programming method based on the above evaluation amount. The fracture quality and fracture area t obtained in the optimal estimation section 3 are replaced with those values previously used in the current estimation section 12, other input data are set to the same values as last time, and the reactor water level is calculated again in the same time interval. , take into account the change in furnace pressure and recalculate the above evaluation value J. If the evaluation value J is larger than the predetermined sufficiently small value C, the optimal estimator 13 estimates the fracture quality and fracture area again.
Similarly, the current state estimating unit 12 repeats 7 simulations.

When such a procedure is repeated and the value of the evaluation quantity J becomes smaller than the predetermined value ε, the fracture quality and fracture area at this time represent the scale of the accident in this case.

Now, the types of process quantities of the simplified model used in the current state estimating section 12 will be briefly described.

The model of the current state estimating unit 12 may use other process variables, such as reactor pressure, reactor water level, feed water flow rate, emergency cooling system flow rate, and enthalpy. However, here, it is modeled as shown in FIG.

In other words, 17 saturated water and 1 saturated steam flow inside the reactor pressure vessel.
It is assumed that the pressure is uniform within the pressure vessel. Also, a heating element 19 is considered to simulate heat generation in the reactor core. The inlet/outlet of the cooling water to the pressure vessel is a saturated steam outlet flow path 20 as shown in the figure.
, a saturated water outlet channel 21, and a cooling water inlet channel 23.

We will model this system by considering the mass balance and energy balance for water and steam.

An example of the results obtained based on this model is shown in FIG.

Now, suppose that a recirculation system pipe rupture occurs with a rupture area of 0, itt" (square feet) and a rupture quality of 0.0. Such an accident is a hypothetical accident, and there is no actual measurement data, but detailed accident An example of analysis using the analysis program is shown by the solid lines in Figure 4.The two solid lines indicate the reactor water level and reactor pressure, respectively.From the time of 1=to, about 2 minutes (T=1
The results of simulation using the data of 25 seconds) according to the above method are shown by broken lines. This result agrees well with the analysis example, and the fracture area in this case is 0.107f.
It can be seen that the fracture area Q, lft'' and fracture quality of 0.0 are well estimated, respectively.

As described above, it is clear that a simplified nuclear reactor model such as the one shown in FIG. 3 can function satisfactorily as long as the fracture area and fracture quality are estimated.

However, the model shown in Figure 3 is extremely simplified compared to an actual nuclear reactor, and it is impossible to predict subsequent long-term phenomena based on such a model. Therefore, the present invention includes a dynamic model that can simulate the dynamic behavior of the internal state of the nuclear reactor with sufficient accuracy.

The above-mentioned fracture size estimation device estimated the fracture area and fracture quality of a pipe rupture accident based on actual measurement data. From this value, it is necessary to determine the position of the pipe in order to set the input values for a dynamic model that can simulate various pipe breaks. Various piping - main steam flow path piping,
There are recirculation flow path piping, water supply flow path piping, and water level instrumentation piping. The relationship between the fracture quality and these piping is determined by the reactor water level,
It can be uniquely determined by the furnace pressure.

For example, when the reactor pressure and reactor water level are maintained sufficiently, the fluid in the recirculation channel piping is a single-phase flow, and in this case, the rupture quality is zero. However, the water supply piping.

The rupture quality of instrumentation piping may also be zero. In this case, by waiting for about 50 to 100 seconds for the furnace pressure to drop, the quality of the water supply piping and instrumentation piping positioned higher in the furnace will be reduced to a certain value (0 x x 1). It turns out. In this case too, the flow in the recirculation channel piping is single-phase. In this way, the fracture position can be uniquely determined from the reactor water level, reactor pressure, and fracture quality.

The database 7 shown in FIG. A decision will be made.

Trigger signal 1 for estimating rupture size as described in Fig. 2
4 is used to set the estimated start time mentioned above, which can be input from the keyboard provided with the CRT. This trigger signal 14 can be used to automate the setting by, for example, capturing an alarm signal of an accident occurrence and simultaneously operating the rupture position estimating device, and operating the same estimating device again after a certain period of time, for example, 50 seconds. It is possible.

Next, a prediction device 5 for predicting the development of phenomena over a long period of time in the event of such an accident will be explained with reference to FIG.

Process signals 8 and 9 are input to the prediction device 5 at regular intervals regardless of whether an accident occurs or not.

The input section 23 is for this purpose, and in this case each signal is subjected to an averaging operation. At the same time, an alarm signal indicating the occurrence of an accident is superimposed on the professional service signal 8.9, and when this signal is activated, the averaging operation is stopped. After several tens of seconds at most, the values of the fracture position and fracture area are inputted as a signal 26. In the model initial value setting section 24, initial values are set based on the steady state process signal and the values of the fracture position and area. As will be described later with reference to FIG. 6, the model of the prediction simulation unit 25 is more complex than the dynamic model for estimating the rupture size described above, and requires initialization settings regarding unmeasured physical quantities. A system of algebraic equations will be prepared for this purpose.

When the initial value setting is performed, the prediction simulation section 2
5, changes in phenomena after an accident are sequentially predicted. At this time, the command to start the prediction and the prediction time can be entered by the operator using the keyboard, but normally the system operates using standard settings.

FIG. 6 shows the scale of the nuclear reactor model used in the prediction simulation section 25. The inside of the reactor is divided into five regions: A non-boiling region in the shroud B saturated water mixing region in the shroud C boiling region outside the shroud D saturated water region outside the shroud E steam region in the dome, and each region stores energy related to liquid and steam. and define the mass balance dynamic equation. Regarding the heat generating part inside the core 27, the fuel rods 27A
The heat transfer from

Various piping, namely recirculation channel piping 28. Water supply piping 29
．． Instrumentation piping 30. The main steam flow path piping 31 is each simulated, as are isolation valves 32 and relief valves 33. In addition, the cooling system is also simulated.

Figures 7 and 8 show the results of simulating 1000 seconds after a recirculation channel rupture accident with a rupture area of 0.1 ft. using such a model. The results of the detailed calculation described above are shown by the solid line.When predicted using this method, even after about 1000 seconds, there is no difference in both the reactor water level and reactor pressure from the detailed calculation (equivalent to actual measured values), and the accuracy is sufficiently guaranteed. It can be seen that the calculation time required for this calculation is about 1/10 of the real time, and sufficient high speed is obtained.

〔Effect of the invention〕

According to the present invention, when predicting changes in the state inside a nuclear reactor after an accident, the scale of the accident is estimated in a short time using a simple dynamic model, and this estimation result and a relatively large-scale model are used to estimate the scale of the accident in a short time. In order to predict the subsequent progress, highly accurate predictions can be made over a long period of 2 to 3 hours, and predictions can be made at a high speed of about 10 times the real time.

[Brief explanation of the drawing]

Fig. 1 is a block diagram showing the configuration of a prediction device of the present invention that predicts the internal state of a reactor at the time of an accident, Fig. 2 is a block diagram of a rupture scale estimating device, and Fig. 3 is a simplified diagram for estimating the rupture scale. Figure 4 is a diagram showing the rupture scale estimation results, Figure #g5 is a block diagram of the prediction device, and Figure 6 is a more detailed reactor model diagram used for relatively long-term prediction after estimation. , FIG. 7 is a diagram showing the predicted result of the reactor water level, and FIG. 8 is a diagram showing the predicted result of the reactor pressure. DESCRIPTION OF SYMBOLS 1... Nuclear reactor, 2... Input device, 3... Accident scale estimation device, 4... Fracture position estimation device, 5... Prediction device L, 6... Display device, 7...・Database, 8,
9... Process signal, 10... State determination section, 11.
. . . Storage device, 12 . . . Current estimation unit, 13 .
6... Steady state signal, 17... Saturated water, 18...
Saturated steam, 19... Heating element, 20... Saturated steam outlet channel, 21... Saturated water outlet channel, 22... Cooling water inlet channel, 23... Input section, 24... - Model initial value setting section, 25... Prediction simulation section, 26...
Fracture position and area signal, 27...core, 27A...
Fuel rod, 28... Recirculation flow path piping, 29... Water supply piping, 30... Instrumentation piping, 31... Main steam flow path piping,
32... Isolation valve, 33... Relief valve. Agent Patent Attorney Tatsuyuki Unuma $4. Katsai 111 (-Sji Kayaza Kata)

Claims (1)

[Claims] 1. In a method for predicting the internal state of a nuclear reactor over a long period of time at the time of an accident, a simple dynamic model is used based on a relatively small number of parameters among the process signals of the reactor. A high-speed estimation stage in which the scale and accident location are estimated and compared with actual measured data in a short period of time, and the estimated values are updated to match the actual measured data. A reactor at the time of an accident in a nuclear reactor, comprising the step of predicting the state inside the reactor over a long period of time by applying the estimated position and the process signal to a relatively low-simplification dynamic model of the reactor. Internal state prediction method. 2. A device that predicts the internal state of a nuclear reactor during an accident over a long period of time uses a simple dynamic model based on trends in a relatively small number of parameters such as reactor pressure and water level among the reactor process signals. An accident scale estimating device that estimates the scale of an accident by using the system, compares it with actual measured data in a short time, and then updates the estimated value so that both agree to match the actual measured data, and the estimated value of the accident scale estimating device. An accident location estimating device that estimates the location where an accident occurred from predicted data, an accident scale estimating device, and estimated values of the accident scale and accident location estimated by the accident location estimating device and the process signal are relatively simplified. A prediction device for predicting the internal state of a nuclear reactor at the time of an accident, characterized by comprising a prediction device that simulates the internal state of the reactor over a long period of time by applying it to a dynamic model of the reactor with a low Device.