JP2011137701A - Device and method for monitoring reactor power - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide the reactor power monitoring technology which improves monitoring accuracy and reliability of nuclear thermal-hydraulic stability. <P>SOLUTION: A reactor power monitor includes calculation components 44, which individually process nuclear instrumentation signals S output from nuclear instrumentation detectors 31, placed inside a core and calculate a first stability index K1 corresponding to each of the nuclear instrumentation detectors 31; and a statistical processing component 70, into which the first stability index K1 is input from each of the calculation components 44 and which statistically processes the index and a first evaluation component 81 which evaluates the nuclear thermal-hydraulic stability in the core from the results of the statistical processing. <P>COPYRIGHT: (C)2011,JPO&INPIT

Description

  The present invention relates to a power monitoring technique for an operating nuclear reactor.

The boiling water reactor (BWR) can control the output by changing the steam flow rate (void ratio) in the boiling core by changing the core flow rate.
However, depending on the core flow rate and other operating conditions, the neutron flux distribution and flow state in the core may become unstable due to void transport delay in the core and the negative feedback effect due to the negative void reactivity coefficient. It has been known.
As a result of this nuclear thermal hydraulic instability phenomenon, there is a concern that the output and flow rate will vibrate greatly, the heat removal characteristics at the fuel rod surface temperature will deteriorate, and the soundness of the fuel rod cladding will be impaired. .

For this reason, when designing the fuel and core of a boiling water reactor, nuclear thermal hydraulic stability analysis is performed to ensure that such vibration phenomenon does not occur in all expected operating areas. Designed to have
And in the area | region where such nuclear thermal hydraulic stability deteriorates, it sets beforehand so that a driving | running | working may be restrict | limited. Also, depending on the type of nuclear reactor, even if it enters into the operation restricted area, there is a type that has the function of reducing the output by inserting a control rod or the like and leaving the operation restricted area.

On the other hand, from the viewpoint of detection and suppression (Detect and Suppress), the output vibration phenomenon is allowed and the output vibration phenomenon caused by nuclear thermal hydraulic instability is accurately detected, and the vibration before the soundness of the fuel is impaired. There are also many nuclear power plants that suppress this.
For this reason, an output vibration detection algorithm called an OPRM (Oscillation Power Range Monitor) using a dedicated detection signal for detecting an output vibration phenomenon has been proposed (for example, Patent Document 1).

Also, there is a known method that performs principal component analysis of vibrations from multiple nuclear instrumentation signals, extracts independent components with different vibration modes, and evaluates the stability of different vibration modes: core integral stability and region stability. (For example, patent document 2).
And, there is a known method for evaluating nuclear thermal hydraulic stability in consideration of the core average neutron flux measurement value (APRM), the delay corresponding to the heat transfer time constant in the fuel rod, the main steam flow rate measurement value, etc. (For example, Patent Document 3).

Furthermore, a device is known that predicts and analyzes the stability of the reduction ratio and issues an alarm when the sequentially detected core stability exceeds the predicted value (for example, Patent Document 4).
In addition, if the indicators that deteriorate the nuclear thermal hydraulic stability of the boiling water reactor are sequentially calculated from plant information such as core power distribution, core flow rate, core pressure, and feed water temperature, and those indicators exceed the set value A technique for issuing an alarm is known (for example, Patent Document 5).
A method of monitoring the stability based on the neutron flux spatial mode distribution obtained by calculating the core eigenvalue from a number of LPRM signals is known (for example, Patent Document 6).

US Pat. No. 5,555,279 JP 2002-221590 A Japanese Patent Laid-Open No. 2002-181984 JP 2000-314793 A JP 2000-121778 A Japanese Patent Laid-Open No. 11-231089

However, in Patent Document 1, the boiling water reactor generally has a decrease in nuclear thermal hydraulic stability as the size, power density, and burnup increase. Cannot be used for type reactors.
In Patent Document 2 to Patent Document 6, nuclear thermohydraulic stability monitoring accuracy has been improved in the past, as a decrease in the margin of nuclear thermohydraulic stability is inevitable as core power and power density improve. It cannot be pursued more than that.

  An object of the present invention is to solve such a problem, and an object of the present invention is to provide a reactor power monitoring technique that improves the monitoring accuracy and reliability of nuclear thermal hydraulic stability.

  The reactor power monitoring apparatus according to the present invention separately processes a plurality of nuclear instrumentation signals output from a plurality of nuclear instrumentation detectors provided in the core, and performs first stabilization corresponding to each of the nuclear instrumentation detectors. A calculation unit that calculates a sex index, a statistical processing unit that inputs a plurality of the first stability indexes from each of the calculation units and executes a statistical process, and a nuclear thermal hydraulic power in the core from the result of the statistical process And a first evaluation unit that evaluates stability.

  According to the present invention, there is provided a reactor power monitoring technique that improves the monitoring accuracy and reliability of nuclear thermal hydraulic stability.

1 is a longitudinal sectional view showing an embodiment of a nuclear power plant to which an output monitoring apparatus according to the present invention is applied. 1 is a horizontal sectional view showing an embodiment of a nuclear reactor to which a power monitoring device according to the present invention is connected. 1 is a configuration block diagram of a reactor power monitoring apparatus according to the present embodiment. The characteristic graph of the band pass filter applied to this embodiment. Reactor operation line graph. The block diagram of a configuration of an autoregressive analysis unit applied to the present embodiment. A graph of time-series data obtained by passing a nuclear instrumentation signal through a bandpass filter. The partial expansion graph of the time series data of FIG. The graph of the damped vibration signal drawn typically. The frequency spectrum of the stability (reduction ratio) by the density function S (f). The overwriting graph of the 1st stability parameter | index (reduction ratio) in a several nuclear instrumentation detector. The graph which displays the maximum value and average value of reduction ratio. The graph which displays the average value and standard deviation (value normalized by the average value) of the reduction ratio. A graph that displays the rate of change of the standard deviation of the reduction ratio. A graph that displays the average value and standard deviation (value normalized by the average value) of the vibration period. The graph which displays the average value, standard deviation, and standardized standard deviation of an amplitude. The flowchart of the reactor power monitoring method by a peak detection function. The graph which displays the phase difference of other nuclear instrumentation signals with respect to a vibration center. Explanatory drawing of the diagnostic method of nuclear thermal hydraulic instability mode. The flowchart of the reactor power monitoring method by a local monitoring function.

Hereinafter, embodiments of the present invention will be described with reference to the accompanying drawings.
A nuclear power generation system shown in FIG. 1 includes a nuclear reactor 10 that generates steam by heating reactor water by heat generated by nuclear fuel fission, a main pipe 21 that guides the generated steam to a turbine 22, and rotationally driven by the steam. A generator 23 that is coaxially connected to the turbine 22 that converts rotational kinetic energy into electrical energy, a condenser 24 that cools, condenses, and condenses the steam expanded by the work of the turbine 22, A water supply pipe 26 that feeds the liquid by the pump 25 and guides it to the nuclear reactor 10 is formed.

  The feed water returned to the nuclear reactor 10 is heated again as reactor water, and the above-described process is repeated to continuously generate power. And the output monitoring apparatus 30 which monitors the nuclear thermal hydraulic stability in the nuclear reactor 10 is provided so that this electric power generation may be maintained stably.

  The nuclear reactor 10 includes a pressure vessel 11 in which a shroud 15 is fixed in an interior filled with reactor water, a core support plate 17 fixed to the shroud 15, and a core support plate 17 supported by the core support plate 17. Are surrounded by a reactor core 16 and a steam / water separator 13 for separating the reactor water that has passed through the reactor core 16 into a gas-liquid two-phase flow.

Then, as described above, one steam separated from the steam by the steam separator 13 is guided to the main pipe 21 to contribute to power generation, and the other separated water joins with the feed water returned from the feed water pipe 26. The combined reactor water is lowered in a region (downcomer D) sandwiched between the shroud 15 and the pressure vessel 11 by a plurality of recirculation pumps 18 provided in the circumferential direction (only one is omitted in the drawing). To the lower plenum L.
Reactor water guided to the lower plenum L passes through the core 16 again and is heated to become a gas-liquid two-phase flow and reaches the upper plenum U. The reached gas-liquid two-phase flow is again guided to the steam separator 13 and the above-described process is repeated.

  As shown in the horizontal sectional view of FIG. 2, the reactor core 16 absorbs neutrons associated with the fission reaction and outputs a rectangular tube-shaped channel box 33 in which a large number of fuel rods (not shown) are accommodated. The control rod 32 to be controlled and the nuclear instrumentation detectors 31a, 31b, 31c, 31d (FIG. 1) for detecting the neutrons are supported, and the upper and lower ends are fixed to the upper lattice plate 14 and the core support plate 17, respectively. A large number of mounting tubes 31 are arranged.

This instrumentation tube 31 is installed in a ratio of about one in 16 channel boxes 33. For example, in an improved boiling water reactor having 872 fuel assemblies, 52 instrumentation tubes 31 are provided. Yes.
The nuclear instrumentation detectors 31a, 31b, 31c, and 31d provided at four locations in the vertical direction of the instrumentation pipe 31 are respectively A level, B level, C level, and D level according to the height position from below. being called. Then, the reactor water circulating inside the reactor core 16 flows in from the A level, is heated by the fuel, starts boiling, and sequentially reaches the B level, the C level, and the D level while changing the water / steam two-phase state. .

Nuclear thermal hydraulic stability is greatly affected by pressure propagation in this two-phase state of water and steam.
That is, due to the delay in the pressure propagation of the reactor water from the bottom to the top of the core 16, the two-phase state (water / steam ratio) fluctuates, and the responses of the nuclear instrumentation detectors 31 a, 31 b, 31 c and 31 d are delayed, A phase difference is generated in each of the nuclear instrumentation signals S (Sa, Sb, Sc, Sd) detected from the B level, C level, and D level.

Since the phase difference of the output oscillation in the reactor water flow direction has an action of canceling the response of the nuclear instrumentation signals S to each other, it is possible to evaluate a plurality of nuclear instrumentation signals S at the same height level. It is preferable from the viewpoint of accuracy and reliability of monitoring of hydrodynamic stability.
Further, it is not necessary to monitor the stability from the A level to the D level, and the nuclear thermal hydraulic stability is evaluated at the B level, which is generally said to have the highest sensitivity of the stability monitoring.

As shown in FIG. 2, the output monitoring device 30 individually processes a plurality of nuclear instrumentation signals S output from the nuclear instrumentation detector 31 at the same height level in the vertical direction in the core 16. The part 40 is provided (in FIG. 2, the flow of the nuclear instrumentation signal S in some of the nuclear instrumentation detectors 31 is indicated by a broken line, and the others are omitted).
As shown in FIG. 3, each signal processing unit 40 corresponds to an A / D converter 41 that digitally converts an analog nuclear instrumentation signal S, a bandpass filter 42, and an individual nuclear instrumentation detector 31. The calculation unit 44 calculates the first stability index K1 and the peak detection unit 60.

  Then, the plurality of first stability indices K1 output from the plurality of signal processing units 40 are input to the statistical processing unit 70 as shown in FIG. 2, and statistical processing is executed. Further, the result of statistical processing in the statistical processing unit 70 is output to the first evaluation unit 81, and the nuclear thermal hydraulic stability in the core 16 is evaluated.

The bandpass filter 42 has a bandpass characteristic as shown in FIG. 4 and passes a frequency band related to nuclear thermal hydraulic stability in the nuclear instrumentation signal S.
When the nuclear instrumentation signal S passes through the band pass filter 42, it becomes time series data Xt as shown in FIG. (FIG. 8 is an enlarged representation by changing the display scale of the time series data Xt)
Here, the frequency of the nuclear instrumentation signal S is proportional to the reciprocal of the time during which the two-phase flow passes through the core 16, and typically has a vibration frequency band of about 0.3 to 0.6 Hz.
Therefore, a frequency band higher than this frequency band is an inherent noise component not related to the evaluation of nuclear thermal hydraulic stability, or a peristaltic component due to mechanical vibration derived from the structure, and thus needs to be excluded. In order to remove such a high-frequency component, the low-pass filter having a cutoff frequency of about 1 Hz is passed.

  In addition, the frequency band lower than the frequency band of output vibration is a trend component that fluctuates slowly due to, for example, plant control, and this low frequency component reduces the calculation accuracy of later autoregressive analysis and peak detection. The filter is removed by passing through a high-pass filter having a cutoff frequency of about 0.1 to 0.2 Hz. This is a process corresponding to normalization because a stationary component is also included. As such a bandpass filter 42, a digital filter such as Butterworth can be applied.

Here, with reference to the operation line of the nuclear reactor 10 shown in FIG. 5, a method of setting the cutoff frequency fcmin of the high-pass filter and the cutoff frequency fcmax of the low-pass filter will be described.
The nuclear reactor 10 is operated at the core power and the core flow rate on the operation line. The operation region where the evaluation of nuclear thermal hydraulic stability is necessary is a low core flow rate / high core power output region, specifically, a region R surrounded by a broken line in FIG.

The frequency of the output vibration is generally proportional to the core output and inversely proportional to the core flow rate. That is, since the left arrow in FIG. 5 indicates the lowest vibration frequency fmin and the right arrow indicates the maximum vibration frequency fmax, the cut-off frequencies fcmin and fcmax are set based on both values.
However, it may be set as shown in the following equation with a slight margin. Here, ε is a constant, for example, a value of about 0.05 is set.

fcmax = fmax + ε
fcmin = fmin−ε

There are two methods for obtaining the minimum frequency fmin and the maximum frequency fmax: a method using actual machine data and a method obtained by analysis.
As a method of using the former actual machine data, statistical processing is performed from the average value signal (APRM) of the nuclear instrumentation signal (S) when the operation state of the actual machine is at a position corresponding to the arrow indicated in FIG. To estimate the vibration frequency.
However, if the core is extremely stable, the vibration frequency may not be estimated with sufficient accuracy. In that case, the vibration frequency can be accurately estimated by estimating the transfer function from the control rod injection reactivity to the neutron flux response by operating the control rod randomly or periodically.

  As a method to obtain by the latter analysis, the nuclear thermal hydraulic stability under the operating conditions indicated by the arrows in FIG. 5 is evaluated by using a stability analysis code that physically models nuclear thermal hydraulic stability of a nuclear reactor. Thus, the vibration frequency can be estimated.

Returning to FIG. 3, the description will be continued.
The calculation unit 44 calculates a reduction ratio, a vibration cycle, and an amplitude related to the first stability index K1 from the time series data Xt, and a reduction ratio calculator 51 and a vibration cycle calculator that are the autoregressive analysis unit 50. 52 and an amplitude calculator 53.
Similarly, the phase difference calculation unit 84 that calculates the phase difference related to the first stability index K1 processes the two time series data Xt to be compared, and thus is provided outside the respective signal processing units 40. ing.

  Here, the attenuation ratio, the vibration period, and the amplitude are defined using the damped vibration signal schematically illustrated in FIG. Assuming that the peak values appear in order as X1, X2, X3, X4,..., And the times when these peaks appear are T1, T2, T3, T4,. And the amplitude are defined as follows:

Reduction ratio = (X3−X4) / (X1−X2)
Vibration period = (T3-T1) or (T4-T2)
Amplitude = (X3-X4) or (X1-X2)
The phase difference is defined by an angle in which a time difference between Tn between a plurality of signals is 360 degrees as one cycle.

The calculation in the calculation unit 44 (FIG. 3) is specifically performed based on autoregressive analysis using an impulse response, a spectral density function, or a transfer function.
As schematically illustrated in FIG. 9, when the reduction ratio, vibration period, amplitude, and the like are calculated from the definition itself, the values rarely vary and become constant values. Because it is a problem.
Therefore, the dynamic response characteristics of the time series signal Xt are estimated by the autoregressive method to obtain a highly reliable stability index. Autoregressive analysis is a technique for linearly estimating the time series signal Xt from past values, and is widely used as a dynamic modeling technique.

In the autoregressive analysis, as shown in FIG. 6, the time series data Xt is linearly fitted to the equation (1) in FIG. 6, and the autoregressive coefficient ak in FIG. 6 (2) is estimated. Here, et is Gaussian noise that is out of the fit. Several known techniques are widely used for algorithms for estimating such autoregressive processes efficiently.
The autoregressive coefficient ak reflects a time-series temporal response characteristic. By using this coefficient ak, it is possible to obtain the reduction ratio, the vibration period, and the phase difference.

  The reduction ratio and vibration period based on the impulse response are averaged from the autoregressive coefficient ak as shown in FIG. 6 (4) through the equation (3) and used as an estimated value with reduced variation.

The reduction ratio and vibration period based on the spectral density function can be obtained as follows. First, using the delay operator z ⁻¹ (x _t−1 = z ⁻¹ x _t ) of discrete (digital) time, the spectral density function represented by the equations of FIG. 6 (5) to FIG. 6 (6). S (f) is obtained. The numerator in FIG. 6 (6) is the variance of Gaussian noise et.

Using this spectral density function S (f), a graph showing the frequency spectrum of stability (reduction ratio) as shown in FIG. 10 is obtained.
The upper spectrum in FIG. 10 shows a stable state (attenuation ratio = 0.2), and the lower spectrum shows the time of oscillation (the attenuation ratio exceeds 1). It can be seen that the more stable the peak of the spectrum tends to be sharp.
Therefore, the reduction ratio can be estimated from the sharpness of the spectrum peak.
As an estimation method, the spectrum is fitted to a quadratic vibration equation including an attenuation constant, the attenuation constant is obtained, and the attenuation constant is converted into a reduction ratio, or the peak height is halved. There is a method of estimating the attenuation constant from the peak width (half width value).
Further, the vibration period is obtained from the reciprocal of the frequency corresponding to the spectrum peak.

The reduction ratio and the vibration period based on the transfer function can be obtained as follows.
6 (7) expressed by the reciprocal number of FIG. 6 (5) is a transfer function from Gaussian noise to time series, and stability information is included in this transfer function.
The zero of A (Z ^-1 ) is the pole of the transfer function, and the stability is determined by the positional relationship of this pole in the complex plane, so by estimating the transfer function pole, the reduction ratio and vibration frequency are Can be estimated.
Here, if the transfer function pole is defined as (P _R , P _I ) as shown in FIG. 6 (9), the reduction ratio and vibration period based on the transfer function are as shown in FIG. 6 (10). Desired. Here, Δt is a sampling period of the time series signal Xt.

In the reduction ratio calculator 51 and the vibration period calculator 52 in the autoregressive analysis unit 50 (FIG. 3), the reduction ratio and the vibration period are calculated by using one of the above three methods. .
The amplitude calculator 53 approximately calculates from the standard deviation of the time series signal Xt to obtain the amplitude. The time series signal Xt used in this case may be the same as the data length used for autoregressive analysis.
The phase difference calculator 84 calculates the phase difference from the cross spectral density of the two time series signals Xt. This cross spectrum can be obtained from a Fourier transform of a cross-correlation function between two signals, an autoregressive analysis of two variables, or the like.

The statistical processing unit 70 performs statistical processing by inputting a plurality of first stability indices K1 from the respective calculation units 44 included in the plurality of signal processing units 40 (FIG. 2). As shown in the figure, the standard deviation calculation unit 71, the average value calculation unit 72, and the maximum value selection unit 73 are configured.
The statistical processing unit 70 includes a reference value set in advance for the first stability index K1 derived from the group of nuclear instrumentation detectors 31 (see FIG. 2) at the same height level in the vertical direction of the core 16. The comparison judgment is carried out to make a judgment on the stability margin of nuclear thermal hydraulic stability.

  Here, the process in which the output vibration grows is examined with reference to the graph of the time series data Xt in FIG. The vicinity of (I) in FIG. 7 is a stable region in which the amplitude of the nuclear instrumentation signal S is small and is regarded as a stationary noise level. This is because the amplitude starts to increase from around (II), and destabilization has begun, completely destabilized near (III), and the amplitude grows rapidly. Become. In this case, it is necessary to reliably detect such instability in the vicinity of (II) to (III) before the state of (IV).

  FIG. 11 displays the overwriting result of the reduction ratio (first stability index K1) output from each calculation unit 44 (FIG. 3) connected to a plurality of different nuclear instrumentation detectors 31 (FIG. 2). FIG. 12 shows the maximum value and the average value of the reduction ratio. According to FIG. 11, the reduction ratio is recognized to vary by the output nuclear instrumentation detector 31, but this variation decreases with destabilization and eventually converges to a certain value. I can see that

FIG. 13 shows the change of the average value of the reduction ratio and its standard deviation (however, the value normalized by the average value), and FIG. 14 shows the change rate of the standard deviation of the reduction ratio.
Although the standard deviation varies in a stable state, the fluctuation starts to decrease near the beginning of destabilization (II), and the fluctuation of the reduction ratio also decreases sharply from the vicinity of (III) where the output oscillation starts to grow rapidly. It can be seen that it starts asymptotic to a certain value.

  FIG. 15 shows changes in the average value of the vibration period (first stability index K1) and its standard deviation (absolute value). According to FIG. 15, the same tendency as the reduction ratio can be recognized in the vibration period.

This is because the system-specific instability becomes obvious due to system instability, and macro vibration pull-in phenomenon occurs, and the discrete vibration of each fuel assembly shifts to macro vibration of the entire core. It represents that. In other words, in the state where the nuclear thermal hydraulic stability of the entire core has become apparent, the reduction ratio and the resonance frequency (vibration period), which are stability indicators, are aligned as a whole. It is possible to determine the instability due to convergence.
The fluctuation itself is smaller in the vibration period (FIG. 15), but the reduction ratio is decreasing more rapidly (see FIG. 13). The vibration period is the absolute value of the fluctuation, and the reduction ratio is It can be seen that monitoring the variation rate of variation is effective for monitoring reliability.

FIG. 16 shows the average value and standard deviation (standard value in absolute value and average value) of the amplitude (first stability index K1). The amplitude increases rapidly due to destabilization, but converges to a limit cycle with a constant amplitude based on nonlinearity. In the example of FIG. 16, the vibration is suppressed before convergence to the limit cycle.
In the case of amplitude, the average value increases rapidly with destabilization, but the variation does not decrease and increases with an increase in average value. This is because the amplitude is not necessarily constant in the core, and the amplitude varies depending on the stability of individual fuel assemblies and the feedback of neutron fluctuations. However, it can be seen that the standard deviation of the normalized amplitude has converged to a constant value.

Returning to FIG. 3, the description will be continued.
The first evaluation unit 81 evaluates the nuclear thermal hydraulic stability through monitoring of the convergence state of variations in the results of statistical processing of the plurality of first stability indices K1 input from the statistical processing unit 70 in this way.
Comparing the convergence situation of the first stability index K1 (FIGS. 11 to 16) with the margin of nuclear thermal hydraulic stability in the time-series data Xt (FIG. 7), it is stable at the earliest time. It can be seen that the standard deviation of the vibration period follows the change in sex. Next, the fluctuation of the amplitude follows, and thereafter, the variation in the reduction ratio rapidly decreases, and finally the average value of the amplitude increases rapidly.

The first evaluation unit 81 evaluates the nuclear thermal hydraulic stability in the reactor core 16 by monitoring the fluctuations of these indices at each stage, thereby closely monitoring the margin of nuclear thermal hydraulic stability.
First, the reduction ratio, but the average value itself is not a very important monitoring parameter here, and the maximum value of the reduction ratio is monitored. Therefore, the standardized standard deviation reduction rate of the reduction ratio is monitored. It is determined that the stability margin has deteriorated when the absolute value of the decrease rate is equal to or less than the set value and continues for a time equal to or greater than the set value, or when the standard deviation value itself falls below the set value.

  In the example of the rate of change of the standardized standard deviation of the reduction ratio shown in FIG. 14, the rate of change continues to become a negative value from 500 seconds to 600 seconds, and from 700 seconds to 900 seconds. It is a section. In the section from 500 seconds to 600 seconds, the value of the standard deviation itself is still larger than 0.15, whereas in the section from 700 seconds to 900 seconds, the value is 0.05 or less. It can be judged that the margin is smaller in the section from 700 seconds to 900 seconds.

  As can be seen from FIG. 13, when the stability begins to deteriorate, the standard deviation of the vibration period decreases almost monotonously. Therefore, depending on the set value, it is possible to set what judgment and what operation is performed in which section.

  As shown in FIG. 16, the amplitude tends to increase suddenly when the stability margin is lost. Therefore, the amplitude is monitored with its maximum value, and the vibration suppressing device 33 (FIG. 3) for ensuring fuel integrity. ) Can be a criterion for starting. At the stage where the standardized standard deviation converges to a substantially constant value, it corresponds to the point in time when the unstable state is saturated and vibration starts to grow, so it can serve as a reference for the preparation stage of vibration suppression (starting of the alarm device 32 or the like). .

Returning to FIG. 3, the description will be continued.
The peak detector 60 includes an amplitude detector 61 and a vibration period detector 62, and derives a second stability index K2 having a higher response than the first stability index K1 from the nuclear instrumentation signal S.
This is because the first stability index K1 derived by autoregressive analysis requires a certain length of data and requires time to obtain a calculation result, and the problem of response delay cannot be avoided.
Therefore, the peak detector 60 is provided to detect the peak of the time series signal Xt and obtain the second stability index K2 such as amplitude and vibration period in a short time.
The first stability index K1 is suitable for monitoring the state transition of the time series signal Xt, and the second stability index K2 is suitable for monitoring a sudden state change of the time series signal Xt.

A method for detecting the amplitude and vibration period according to the second stability index K2 will be described with reference to FIG.
Here, the A value is the amplitude value A obtained in advance by the calculation unit 44 (FIG. 3). This amplitude value A is an average amplitude at the time of tracing back the time for the data length used in the autoregressive analysis. A threshold value pA for peak detection is set by multiplying the amplitude value A by one or more coefficients p. This threshold value pA is assumed to change nonlinearly with respect to the amplitude value A. When the amplitude value A is small, the threshold value pA is large. It will be 1.

The amplitude related to the second stability index K2 includes the peak top and peak bottom of the nuclear instrumentation signal S exceeding the region (region of -pA to pA) obtained by multiplying the amplitude related to the first stability index K1 by one or more coefficients. The oscillation period according to the second stability index K2 is derived from the interval between adjacent peak tops or the interval between adjacent peak bottoms of the nuclear instrumentation signal S exceeding the region.
That is, if the time series data Xt exceeds the threshold value pA, the next peak should come at a time position within half of the vibration period estimated by autoregressive analysis. Therefore, the next peak is searched for at a time position within half of the vibration period from the time point when the time series data Xt exceeds the threshold value pA.
If the next peak is found, the time difference from the previous peak is used as a new vibration cycle, and is used for the next peak detection.

  The second evaluation unit 82 evaluates nuclear thermal hydraulic stability in the core 16 from the second stability index K2. For example, if the amplitude related to the second stability index K2 is detected twice continuously, exceeding the threshold value B in FIG. 8, the alarm device 34 informs the operator, and if detected continuously 10 times. The alarm device 35 is operated to automatically insert a control rod to reduce the output and suppress the output vibration.

  As described above, when the first evaluation unit 81 evaluates the destabilization of nuclear thermal hydraulic stability, the evaluation by the first evaluation unit 81 is stopped and the evaluation is switched to the high response evaluation by the second evaluation unit 82. Therefore, it is possible to detect the instability of the nuclear thermal hydraulic power state of the core with high reliability and with sufficient margin before the fuel soundness due to the output vibration is threatened. it can.

Next, referring to the flowchart shown in FIG. 17 (see FIG. 3 as appropriate), the operation of the reactor power monitoring method when the peak detection function is used will be described.
First, time series data Xt obtained by individually processing a plurality of nuclear instrumentation signals S output from a plurality of nuclear instrumentation detectors 31 in the group shown in FIG. 2 is input to the calculation unit 44 (FIG. 3) (S11). The first stability index K1 corresponding to each nuclear instrumentation detector 31 is calculated (S12). Next, a plurality of first stability indexes K1 output from the respective calculation units 44 are input to the statistical processing unit 70, and statistical processing is executed (S13).

Then, in the first evaluation unit 81, the nuclear thermal hydraulic stability is evaluated from the result of the statistical processing, and it is determined whether or not the stability margin is deteriorated (S14).
If it is determined that the stability margin has not deteriorated (S14: No), the nuclear thermal hydraulic stability is continuously evaluated based on the first stability index K1.

  On the other hand, if it is determined that the stability margin has deteriorated (S14: Yes), the time series data Xt is transmitted to the peak detector 60, and peak detection for high response evaluation is started (S15). Then, an average value A of past amplitudes is acquired from the first stability index K1 calculated in S12 (S16). Next, the first threshold value pA and the second threshold value B are set (S17), and when the time series data Xt exceeds the first threshold value pA (S18: Yes), the peak is detected (S19), and the second stability. The amplitude and vibration period related to the index K2 are detected (S20).

  When the amplitude related to the second stability index K2 exceeds the second threshold B (S21: Yes), it is evaluated that the nuclear thermal hydraulic stability has become unstable, and a warning warning is issued (S22). The safety device is activated (S23). In the case of (S18: No) (S21: No), the peak detection function is continued (S15).

Next, returning to FIG. 3, the local monitoring unit 85 will be described.
For example, when the reduction ratio according to the first stability index K1 shown in FIG. 12 is significantly larger than the index values of other nuclear instrumentation detectors, the nuclear instrumentation detector 31 is concerned. It is considered that local vibration phenomenon in the vicinity of the core region, or region vibration around that region has occurred or is likely to occur.

The local monitoring unit 85 has a function of monitoring the area with special attention.
That is, when the first evaluation unit 81 evaluates the destabilization of nuclear thermal hydraulic stability, the local monitoring unit 85 significantly destabilizes the nuclear instrumentation detector 31 among the plurality of first stability indices K1. And focus monitoring the surrounding local area.
Then, the local monitoring unit 85 monitors at least one of the reduction ratio, amplitude, and phase difference of the nuclear instrumentation signal S in the region expanded from the set local region, and performs local vibration, region vibration, and core integrated vibration. The nuclear thermal hydraulic instability mode is estimated.

  For example, monitoring the attenuation ratio of the detector adjacent to the detector that gives the maximum attenuation ratio, and if those values are significantly different from the detector showing the maximum attenuation ratio, the stability in the local region Since the condition has deteriorated, the amplitude is carefully monitored, and if the amplitude exceeds the limit value, an operation to suppress vibration is started. Here, as the limit value, an output vibration amplitude that may impair fuel soundness is analytically obtained in advance, and a value with a margin is set for the amplitude.

  Here, when the adjacent range is gradually expanded from the central region of the reduction ratio maximum shown in FIG. 19 as shown by the solid line to the broken line, and enlarged to a region where a significant difference in the reduction ratio or amplitude occurs. The range of output vibration can be estimated. When the range extends over the entire core, the nuclear thermal hydraulic instability mode can be determined as the instability of the core integral is the region instability when staying on one side of the core.

Further, as shown in the phase difference response of FIG. 18, the unstable mode can be similarly estimated by sequentially evaluating the phase difference with the nuclear instrumentation signal S at the estimated center point of vibration.
At the stage where there is a margin of stability, the phase difference varies with a value of about several tens of degrees, but at the stage where destabilization starts around 600 seconds, the position of the detector signal on the left and lower left is The phase difference rapidly increases and the phase difference is approximately 180 degrees, that is, in an antiphase state, but the phase difference with the lower detector remains at about 30 degrees or less. From this, it can be seen that in this example, region vibration or region instability having the opposite phase to the half core region in the left direction of the core from the center point of vibration occurs.

The above procedure is shown in FIG.
Assuming that the vibration center point is the detector 31A, first, the four most adjacent detectors 31B, 31C, 31D, and 31E in a solid line circle shape are monitored. If it is determined that the vibration has spread to this point, the monitoring range is then expanded to the outer circumference of the broken line or to the adjacent region. Such processing is repeated to estimate the spatial extent of vibration.

In addition, as shown in FIG. 19, the phase difference between the nuclear instrumentation detector 31A and the detectors 31F, 31G, and 31H farthest in the core center direction is monitored, and if the phase difference is about several tens of degrees. It can be determined that the instability of the core in the same phase in the reactor core and the instability of the region having a significant phase difference can be determined as long as the phase difference is larger than 100 degrees.
If there is no clear phase difference and only the detector 31A protrudes with respect to the amplitude and the reduction ratio, it can be determined that there is local instability. In the case of local instability, for example, the signal response of the five detectors 31A to 31D including the detectors on the circumference of the solid line is intensively monitored to ensure an output oscillation phenomenon that impairs fuel integrity. It is possible to avoid it.

Next, referring to the flowchart shown in FIG. 20 (see FIG. 3 as appropriate), the operation of the reactor power monitoring method when the local monitoring function is used will be described.
Note that S31 to S34 in FIG. 20 are the same as S11 to S14 in FIG.
When the local monitoring function operates (S35), the detector 31 having the maximum reduction ratio is identified and the detector 31A is estimated as the vibration center point (S36). Next, the peripheral detectors 31B, 31C, 31D, and 31E are observed (S37), and the presence or absence of a significant difference in the reduction ratio signal is confirmed (S38).
If it is determined that there is no significant difference (S38: No), it is determined that the mode is a local instability mode (S42), a warning to that effect is made (S43), and the safety device is activated (S44).

  On the other hand, if it is determined that there is a significant difference (S38: Yes), the observation range is expanded (S39), the range of the output vibration is estimated (S40), and the detectors 31F, 31G, 31H at remote positions are detected. Is evaluated (S41). Then, the type of unstable mode is determined from the value of this phase difference (S42), a caution alarm to that effect is made (S43), and the safety device is activated (S44).

  The present invention is not limited to the above-described embodiments, and can be appropriately modified and implemented within the scope of the common technical idea. For example, the output monitoring device 30 is a computer including such an execution unit, and includes a case in which an operation or data processing designated based on a program is executed.

  DESCRIPTION OF SYMBOLS 10 ... Reactor, 11 ... Pressure vessel, 11 ... Reactor, 13 ... Steam separator, 14 ... Upper lattice plate, 15 ... Shroud, 16 ... Core, 17 ... Core support plate, 18 ... Recirculation pump, 21 ... Main pipe, 22 ... turbine, 23 ... generator, 24 ... condenser, 26 ... water supply pipe, 30 ... output monitoring device, 31 ... instrumentation pipe, 31a, 31b, 31c, 31c ... nuclear instrumentation detector, 32 ... Control rod 33 ... Channel box 34 ... Alarm device 35 ... Output suppression device 40 ... Signal processing unit 41 ... A / D converter 42 ... Band pass filter 44 ... Calculation unit 50 ... Auto regression analysis unit , 51 ... Reduction ratio calculator, 52 ... Vibration period calculator, 53 ... Amplitude calculator, 60 ... Peak detector (detector), 61 ... Amplitude detector, 62 ... Vibration period detector, 70 ... Statistical processor , 71 ... Standard deviation calculator, 72 ... Average value calculator 73: Maximum value selection unit, 81 ... First evaluation unit, 82 ... Second evaluation unit, 84 ... Phase difference calculation unit, 85 ... Local monitoring unit, K1 ... First stability index, K2 ... Second stability index, S: Nuclear instrumentation signal, Xt: Time series data.

Claims (14)

A calculation unit that individually processes a plurality of nuclear instrumentation signals output from a plurality of nuclear instrumentation detectors provided in the core and calculates a first stability index corresponding to each of the nuclear instrumentation detectors;
A statistical processing unit that performs a statistical process by inputting a plurality of the first stability indices from each of the calculation units;
A reactor power monitoring apparatus, comprising: a first evaluation unit that evaluates nuclear thermal hydraulic stability in the core from the result of the statistical processing.   2. The reactor output according to claim 1, wherein the plurality of nuclear instrumentation signals to be processed are outputted from the nuclear instrumentation detector at the same vertical height level in the core. Monitoring device.   The reactor power monitoring apparatus according to claim 1, further comprising: a band-pass filter that allows a frequency band related to the nuclear thermal hydraulic stability in the nuclear instrumentation signal to pass therethrough.   The reactor output according to any one of claims 1 to 3, wherein the first stability index is at least one of a reduction ratio, a vibration period, an amplitude, and a phase difference. Monitoring device.   5. The nuclear reactor according to claim 1, wherein the calculation in the calculation unit is based on an autoregressive analysis using an impulse response, a spectral density function, or a transfer function. Output monitoring device.   The said 1st evaluation part evaluates the said nuclear thermal hydraulic stability through the monitoring of the convergence state of the dispersion | variation in the result of the said statistical process in these several 1st stability parameter | index, The Claim 1 characterized by the above-mentioned. The reactor power monitoring apparatus according to any one of claims 5 to 5.   The reactor power monitoring apparatus according to claim 6, wherein the target of monitoring the convergence state is any one of a reduction ratio, a vibration period, and an amplitude, or a combination thereof. A detector for deriving a second stability index from the nuclear instrumentation signal with a higher response than the first stability index;
The reactor according to any one of claims 1 to 7, further comprising: a second evaluation unit that evaluates nuclear thermal hydraulic stability in the core from the second stability index. Output monitoring device. The detector is
Deriving the amplitude related to the second stability index from the interval between the peak top and the peak bottom of the nuclear instrumentation signal exceeding the region obtained by multiplying the amplitude related to the first stability index by one or more coefficients,
The nuclear reactor according to claim 8, wherein a vibration period related to the second stability index is derived from an interval between adjacent peak tops or an interval between adjacent peak bottoms of a nuclear instrumentation signal exceeding the region. Output monitoring device. When the first evaluation unit evaluates the destabilization of the nuclear thermal hydraulic stability,
The reactor power monitoring apparatus according to claim 8 or 9, wherein the evaluation by the first evaluation unit is stopped and the evaluation is switched to the evaluation by the second evaluation unit. When the second evaluation unit evaluates the destabilization of the nuclear thermal hydraulic stability,
An alarm device to warn the operator to that effect,
The reactor power monitoring apparatus according to any one of claims 8 to 10, further comprising: an output suppressing unit that suppresses the reactor output. When the first evaluation unit evaluates the destabilization of the nuclear thermal hydraulic stability,
The nuclear instrumentation detector that is significantly destabilized from the plurality of first stability indicators is identified, and a local region for performing focus monitoring is set around the destabilized nuclear instrumentation detector. The reactor power monitoring apparatus according to claim 1, further comprising a monitoring unit.   The local monitoring unit monitors at least one of a reduction ratio, an amplitude, and a phase difference of a nuclear instrumentation signal in a region expanded from the set local region, and performs nuclear heat such as local vibration, region vibration, and core integrated vibration. The reactor power monitoring apparatus according to claim 12, wherein a hydraulic instability mode is estimated. A calculation step of individually processing a plurality of nuclear instrumentation signals output from a plurality of nuclear instrumentation detectors provided in the core to calculate a first stability index corresponding to each of the nuclear instrumentation detectors;
A statistical processing step of performing statistical processing by inputting a plurality of the first stability indices from each of the calculation units;
And a first evaluation step for evaluating nuclear thermal hydraulic stability in the core from the result of the statistical processing.

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