JPH0810265B2 - REACTOR REACTIVITY MONITORING METHOD AND DEVICE - Google Patents

Description

TECHNICAL FIELD OF THE INVENTION The present invention relates to a reactor reactivity monitoring method and apparatus using an output signal of a detector in a reactor.

[Technical background of the invention and its problems] In a boiling water reactor, an in-reactor detector is used as a nuclear instrumentation system for measuring the neutron flux level. Depending on the level, a neutron source region monitor, an intermediate region monitor, and The power range monitor is used properly, but the operator has empirically determined the criticality of the core based on the time change of the output of the neutron source range monitor.

The reactivity ρ is obtained by the following equation by measuring the furnace period T from the time change of the output.

Where β is the total delayed neutron production rate, βi is the delayed neutron production rate of the i-th group, λi is the delayed neutron preceding nuclear lifetime of the i-th group, l is the neutron lifetime, and ρ / β is β The dollar reactivity in units. When operating a nuclear reactor, a conversion table for T and ρ / β is usually prepared in advance and used.

FIG. 11 shows a conventional reactor core reactivity monitoring apparatus. In the figure, reference numeral 1 is a reactor in which the reactivity of the reactor core 2 is controlled by a control rod 100. The neutron flux level of the core 2 is detected by the in-core detector 3 of the neutron source area monitor, and its pulse signal is passed through the amplifier 8 and then converted into the logarithmic count rate signal 20 by the logarithmic conversion circuit 9 to monitor the neutron source area. Is displayed on the indicator 101, the recorder 102, and the CRT screen 7. Also, the logarithmic count rate signal 20 is output as a furnace period signal by the differentiating circuit 103, and the period indicator 10
4 and CRT screen 7 are displayed. In such a conventional device, the reactor period can be measured as 1.44 times the doubling time of the counting rate of the neutron source region monitor, but the time change of the counting rate after interrupting the operation of the control rod 100 is beyond the critical value. In some cases, as shown in FIG. 12, the rate of rise did not become constant until 80 seconds or more passed after the control rod withdrawal operation was interrupted, and it took several minutes to measure the furnace period. In the subcritical state, the corresponding reactor period should be negative, but before the logarithmic count rate decreases, as shown in Fig. 13, the output settles at a constant count rate determined by the external neutron source. Moreover, the measurement of the furnace period was impossible.

Thus, with the conventional reactivity calculation method, it is possible to measure only when the required furnace period is positive, and even with a positive furnace period, it takes time to measure and it is not possible to immediately calculate the reactivity. It was Furthermore, the count rate signal from the neutron source area monitor usually contains fluctuations,
There was uncertainty in the measurement of the furnace period.

[Object of the Invention] The present invention has been made in consideration of such a point, and from the large negative reactivity when the control rods are fully inserted into the core without measuring the reactor period, a positive reaction exceeding the criticality is obtained. It is intended to provide a reactor reactivity monitoring method and apparatus capable of presenting the reactivity of the core to an operator in real time.

[Summary of the Invention] That is, according to the present invention, the output signals from a plurality of in-reactor detectors for measuring the neutron flux level are sampled, and the output signals and detectors of channels in which the detectors are bypassed or malfunctioning are all detected. A neutron density measuring device that outputs an average neutron density signal by averaging except the output signal of the channel not at the insertion position and the output signal having a particularly large deviation,
A time-series data recording device that records the latest neutron density signal from this neutron density measuring device as time-series data, and the latest neutrons by inputting the data group of the neutron density signal from this time-series data recording device. The most probable value of the density signal is calculated by the function approximation, and the core reactivity derived from the eigenvalue of the external neutron source value from the eigenvalue by the core calculation and the neutron density signal corrected by the function approximation are used for the one-point reactor approximate dynamic characteristic equation. The neutron source value and the neutron density signal corrected by the above function approximation are applied to the one-point reactor approximate dynamic characteristic equation to obtain a wide range from the full insertion of the control rod to the supercritical range. And a reactivity calculation device that calculates the reactivity of the core in real time.

Further, the present invention, the output signal from the plurality of in-reactor detectors for measuring the neutron flux level, the output signal of the channel in which the detector is bypassing or malfunctioning, and the output signal of the channel in which the detector is not at all insertion positions. And an averaging process step of calculating an average neutron density signal by performing an averaging process excluding particularly large deviation output signals, and a data accumulating step of accumulating the average neutron density signal periodically calculated as time series data. , The neutron density correction process of calculating the most probable value of the latest neutron density signal from the accumulated time series data of the average neutron density signal by function approximation, and the core calculation based on the diffusion equation when the reactor is in the subcritical steady state. The reactivity of the current core state is calculated from the obtained eigenvalue, and the calculated reactivity and the latest neutron density signal corrected by the neutron density correction process The external neutron source calculation step of calculating the fixed external neutron source value by applying to the single-point reactor approximate dynamic characteristic equation, and the fixed external neutron source value calculated by this external neutron source calculation step And a reactivity calculation step of calculating the reactivity of the core in real time from the latest neutron density signal corrected by the neutron density correction step applied to the reactor reactivity monitoring method.

BEST MODE FOR CARRYING OUT THE INVENTION [Embodiments of the Invention] Hereinafter, the present invention will be described in detail with reference to embodiments shown in the drawings.

FIG. 1 shows an embodiment of the present invention connected to a neutron source area monitor. The reactor reactivity monitoring apparatus of the present invention mainly comprises a plurality of in-reactor detectors 3 installed in a core 2 of a reactor 1,
A neutron density measuring device 4 that inputs the output signal of the in-core detector 3 and outputs an average neutron density signal, and a time series that records the latest constant time neutron density signal from the neutron density measuring device 4 as time series data. A data recorder 5 and a reactivity calculator 6 for inputting a data group of neutron density signals from the time-series data recorder 5 to calculate the reactivity of the core in real time.
And a display device such as a CRT screen 7.

In the figure, reference numerals 3A, 3B, 3C and 3D are 4 of the neutron source area monitor.
The in-core detectors are installed at different positions of the core 2. The output signals from each of these detectors are processed by four independent systems, and are processed in channels A, B, and
These are called C and D signals. The detection signals from the in-reactor detectors 3A, 3B, 3C, 3D are input to the logarithmic conversion circuits 9A, 9B, 9C, 9D via the amplifiers 8A, 8B, 8C, 8D, where the logarithmic calculation rate signal 20
Converted to A, 20B, 20C, 20D. Neutron density measuring device 4
Is composed of a sampling processing mechanism 10 and an averaging processing mechanism 11, and the sampling processing mechanism 10 has a logarithmic count rate signal 20A, 2
0B, 20C, and 20D are input, and the sampling period signal 21 is input.
It is converted into the original counting rate by being sampled at a constant cycle based on. The sampling period signal 21 is defined in the range of 1 to 10 seconds for the reason described below. The output signals 22A, 22B, 22C, 22D from the sampling processing mechanism 10 are input to the averaging processing mechanism 11 together with the bypass signals 23A, 23B, 23C, 23D and the in-core detector insertion state signals 24A, 24B, 24C, 24D, It is processed into the average neutron density signal 25, which is input to the time series data recording device 5 together with the time signal 26. The reactivity calculation device 6 sequentially inputs the latest data from the time-series data recording device 5, calculates the reactivity of the core in real time based on this data, and outputs it to the display device such as the CRT 7.

The operation of the reactor reactivity monitoring device configured as described above will be described below. The detection signals of the in-core detectors 3A, 3B, 3C, 3D that detect the neutron flux level at each position of the core 2 are amplified by amplifiers 8A, 8B, 8C, 8D, and then logarithmic conversion circuits 9A, 9B, After being converted into logarithmic count rate signals 20A, 20B, 20C and 20D in 9C and 9D, they are sampled at a constant cycle for digital processing by the sampling processing mechanism 10 and converted into the original count rate. Then, in the averaging processing mechanism 11, the channel in bypass is excluded based on the bypass signals 23A, 23B, 23C, 23D from the neutron density signals 22A, 22B, 22C, 22D output from the sampling processing mechanism 10, and then the reactor. Inner detector insertion status signal 24A, 2
Based on 4B, 24C, 24D, channels not in all insertion positions are excluded, and finally the neutron density signal of the remaining channels 22
If there is a large deviation among these, the neutron density signal is averaged by excluding it. The average neutron density signal 25 thus processed together with the time signal 26 as time series data of the neutron density signal
Recorded in.

The reactivity calculation device 6 starts the period calculation of the reactivity when the data in the time-series data recording device 5 is accumulated by the number required for the function approximation. 2 and 3 show flowcharts of reactivity cycle calculation. First, the most probable value of the latest neutron density signal is obtained by function approximation. When the core becomes critical or supercritical, the neutron density signal increases according to an exponential function, so a first-order exponential function is suitable as an approximate function. When the function of the latest neutron density signal is set to 0 and the function is obtained by least square fitting, it is expressed by the following equation.

N (t) = exp (a + bt) where a and b are constant terms and first-order coefficients obtained by least-squares fitting, t is time in unit second, and N is
(T) is the most probable value of the neutron density signal at time t. From this, the latest neutron density signal N and the most probable value of the rate of change over time are obtained by the following equation.

N = exp (a) dN / dt = b · exp (a) However, regarding the function approximation of the neutron density signal, if the target time is too short, the number of digital data points will decrease, or the error in the time change of the neutron density signal will occur. Becomes large, and conversely, if the time is too long, the time change of the neutron density signal deviates from the approximate function, so the data from 5 seconds to 1 minute before the latest data should be targeted. In the following calculation, the value corrected by the above function approximation is used as the latest neutron density signal N, and the fluctuation component of the neutron density signal is removed.

Next, the modified neutron density signal N is applied to the following one-point reactor approximate dynamic characteristic equation to calculate the reactivity ρ of the core.

dCi / dt = Nβi / l−λiCi (II) Here, Ci (i = 1 to 6) is the delayed neutron precursor nucleus density of the i-th group, and S is an external neutron source. To calculate the reactivity ρ from the neutron density signal N, the delayed neutron preceding nuclear density
Ci and an external neutron source S are required.

The external neutron source S is confirmed when the reactivity calculation has been previously performed for the target core, but when the reactivity calculation is performed for the first time or when the core combustion further progresses, the external neutron source S If is not determined, the core calculation must be performed by diffusion calculation for the current core state as shown in FIG.

In order to calculate the external neutron source S with high accuracy, it is desirable that the core be in a subcritical state with a large negative reactivity ρ. Normally, this condition is satisfied if calculation is performed for the core with all control rods inserted. If three-dimensional coarse mesh diffusion calculation is performed by a computer as core calculation, the eigenvalue Ke
ff can be obtained, and the reactivity ρ can be obtained from this eigenvalue Keff by the following equation.

ρ = (Keff−1) / Keff (III) Therefore, in the subcritical steady state, the external neutron source S can be calculated by the following equation from the equations (I) and (II).

S = -Nρ / l (IV) At the same time, the initial value of the delayed neutron precursor nuclear density Ci can be calculated according to the following equation.

Ci = Nβi / (λi · l) (V) Even if the external neutron source S has already been calculated, the initial value of the delayed neutron precursor nuclear density Ci is calculated by the formula (V) at the start of the reactivity period calculation. There is a need.

As a method for calculating the external neutron source S, there is also a method as shown in FIG. That is, the reactor water temperature, core average control rod density, core average xenon concentration, and core average burnup are changed as variables of the core state, and core calculation by diffusion calculation is performed in advance to create an eigenvalue table. From the eigenvalue table, the reactor core water temperature, the core mean control rod density, the core mean xenon concentration, and the core mean burnup eigenvalue Keff corresponding to the current core state are calculated by internal and external interpolation. The reactivity ρ is calculated, and then the external neutron source S is calculated by the equation (IV).

Next, after calculating the initial values of the external neutron source S and the delayed neutron preceding nuclear density Ci, the single-point reactor approximate dynamic characteristic equations of the formulas (I) and (II) are solved by the time-difference approximation to obtain the delayed neutrons. Calculate the preceding nuclear density Ci. When the neutron density signal N is obtained at time t, the delayed neutron preceding nucleus density Ci at time (t + Δt) is calculated by the following equation.

Ci (t ＋ △ t) ＝ [(1 ＋ λi △ t / 2)] × [N △ tβi / l ＋ (1－λi △ t / 2) Ci (△ t)] …… (VI) where (VI) equation In order to improve the calculation accuracy of, it is necessary to make Δt as small as possible, and the maximum value of λi is about 3
Since it is (1 / sec), Δt is preferably 0.5 sec or less. Since this value tends to be smaller than that of the sampling cycle signal 21, it is desirable that the sampling cycle be an integral multiple of Δt. However, if Δt is determined by an integer fraction of the sampling period, the calculation accuracy may suddenly deteriorate if the sampling period changes, and if Δt is made too small, the calculation amount will increase. Is set to an integral fraction of 1 second such as 0.2 seconds or 0.25 seconds, and the sampling period 21 is set to 1 second unit,
The number of iterations of equation (VI) is calculated for each reactivity cycle calculation. In this way, the sampling period signal 21 does not necessarily have to be constant. However, if the sampling period becomes too long, the accuracy of the function approximation deteriorates, so the sampling period should be set within the range of 1 to 10 seconds.

If the delayed neutron preceding nuclear density Ci is calculated by the equation (VI),
Since the formula (I) becomes the following formula, the reactivity ρ can be calculated according to the following formula.

In this way, the reactivity calculation device 6 calculates the reactivity cycle. Since the time required for this calculation is less than 1 second, the reactivity of the core can be calculated in substantially real time.

The result of this calculation is displayed on a display device such as the CRT screen 7, but it is easier for the operator to understand the dollar reactivity in β units, so ρ / β is displayed digitally or as a trend on the CRT screen. It

As described above, the reactivity of the core is presented to the operator in real time using the output signal of the neutron source region monitor as an input.
Operators can constantly monitor the nuclear safety of the core,
It is possible to smoothly operate the reactivity by operating the control rod. This greatly improves the working efficiency of the operator especially regarding the control rod operation up to the criticality at the time of reactor startup. FIGS. 4 (a) and 4 (b) show information given to the operator by the present invention until reaching the criticality at the time of reactor startup and an example of the operator's control rod operation using this information. Based on the control rod operation amount and the change in reactivity, the operator contributes to the reactivity per control rod (called control rod value) from about $ 0.2 to 0.3.
Assuming that the information is about dollars, the operator can pull out the control rods one after another without having to worry about the restrictions, provided that there is a margin of one dollar or more before the criticality. You can see how it approaches criticality by changing the reactivity.
It is only necessary to reduce the operation amount near the criticality, and it is not necessary to interrupt the drawing operation until the criticality is confirmed. On the other hand, when the critical approach is performed by the conventional change of the count rate signal, the operator controls the control rod until the increase rate of the count rate signal decreases as shown in FIG. The operation has to be interrupted, so that the control rod operation amount per unit time becomes small as compared with the case of using the present invention, and the time until the criticality is achieved takes that much longer. Also, the burden on the operator is large.

In the reactivity calculation, the method of using the eigenvalue table for determining the external neutron source is illustrated, but the reactivity ρ may be used as a table instead of the eigenvalue table to calculate the reactivity.

Next, an embodiment of the present invention utilizing the output of the intermediate area monitor will be described with reference to FIG. In the figure, the same parts as those in FIG. 1 are designated by the same reference numerals. In the reactor core 2 of the reactor 1, eight in-core detectors 3a, 3 are provided as intermediate region monitors.
b, 3c, 3d, 3e, 3f, 3g, and 3h are installed at different positions, and their output signals are processed by eight independent channels, and channels a, b, c, d, e, and f are respectively processed. , G, h
Called the signal of. Furnace detectors 3a, 3b, 3c, 3d, 3e, 3
Output signals of f, 3g, 3h are amplifiers 8a, 8b, 8c, 8d, 8e, 8
After going through f, 8g, and 8h, range signals 27a, 27b, 27c, 27d, 2
Gain adjusters 12a, 12b, 12 with 7e, 27f, 27g, 27h
c, 12d, 12e, 12f, 12g, 12h level-adjusted, then the level-adjusted output signal is bisected, one of them is the root mean square circuit 13a, 13b, 13c, 13d, 13e, 13f, 13g, 1
3h, the other phase inverter 14a, 14b, 14c, 14d, 14e, 14
After the root mean square circuit 13a, 13b, 13 processed by f, 14g, 14h
It is input to c, 13d, 13e, 13f, 13g and 13h. Intermediate area monitor output signals 28a, 28b, 28c, 2 output from these root mean square circuits 13a, 13b, 13c, 13d, 13e, 13f, 13g, 13h.
8d, 28e, 28f, 28g, 28h are range signals 27a, 27b, 27c, 2
While being displayed on the display device such as the CRT screen 7 together with 7d, 27e, 27f, 27g, and 27h, it is input to the neutron density measuring device 4. Intermediate area monitor output signals 28a, 28b, 28c, 28d, 28
e, 28f, 28g, 28h are range signals 27a, 27b, 27c, 27d, 27
Since the gain of the output signal is adjusted according to e, 27f, 27g, and 27h, it cannot be directly used as a neutron density signal. Therefore, the neutron density measuring device 4 has a neutron density signal conversion processing mechanism 15 in addition to the sampling processing mechanism 10 and the averaging processing mechanism 11.

In the neutron density signal conversion processing mechanism 15, when the neutron source area monitor and the intermediate area monitor are at a usable neutron flux level in an overlapping manner, the logarithmic count rate signals 20A, 20B, 20C, 20D of the neutron source area monitor are created. The averaged neutron density signal 25 is used for adjusting the neutron density signal of the output signal from the intermediate area monitor. That is, the sampling processing mechanism 10 is the logarithmic count rate signal of the neutron source area monitor.
20A, 20B, 20C, 20D and intermediate area monitor output signals 28a, 28
b, 28c, 28d, 28e, 28f, 28g, 28h is sampled by the sampling period signal 21, and the neutron source region monitor outputs the neutron density signals 22A, 22B, 22C, 22D to the averaging processing mechanism 11, while the intermediate region is output. Monitor instruction signal 29a, 29
b, 29c, 29d, 29e, 29f, 29g, 29h are output to the neutron density signal conversion processing mechanism 15. Neutron density signal conversion processing mechanism
15 is an intermediate area monitor instruction signal 29a, 29b, 29c, 29d, 29.
Range signals 27a, 27b, 27 along with e, 29f, 29g, 29h
c, 27d, 27e, 27f, 27g, 27h and bypass signals 23a, 23
b, 23c, 23d, 23e, 23f, 23g, 23h and in-reactor detector insertion state signals 24a, 24b, 24c, 24d, 24e, 24f, 24g, 24h,
The average neutron density signal 25 from the averaging processing mechanism 11 is input and the converted neutron density signal 30 is output. In the time-series data recording device 5, the average neutron density signal 25 from the averaging processing mechanism 11 is selected by the neutron density signal selecting mechanism 16 until the neutron source region monitor reaches the upper limit of use, and the time signal is recorded.
Although it is input together with 26, in the range of use of the intermediate region monitor which is the upper limit of the use of the neutron source region monitor, the reduced neutron density signal 30 is selected and input together with the time signal 26,
Will be recorded. The reactivity calculation device 6 sequentially inputs the latest data from the time series data recording device 5 to calculate the reactivity of the core, and displays the result on a display device such as the CRT screen 7.

Next, the operation of the reactivity monitoring device having the above configuration will be described. In-reactor detectors 3a, 3b, 3c, 3d, 3e of the intermediate range monitor,
The output signals from 3f, 3g, 3h are amplified and level-adjusted, and then the root mean square circuits 13a, 13b, 13c, 13d, 13e, 13f, 13
Intermediate area monitor output signals 28a, 28b, 28c, 28 from g, 13h
Sampling mechanism 10 as d, 28e, 28f, 28g, 28h
Is output to In the sampling processing mechanism 10, the intermediate area monitor output signals 28a, 28b, 28c, 28d, 28e, 28f, 28g, 28h.
Is sampled at a constant cycle, and the intermediate area indication signals 29a, 2
9b, 29c, 29d, 29e, 29f, 29g, while outputting to the neutron density signal conversion processing mechanism 15 as 29h, up to the upper limit of use of the neutron source region monitor log count rate signal 20A, 20B, 20C from the neutron source region monitor , 20D are also sampled at a constant cycle and are output to the averaging processing mechanism 11 as neutron density signals 22A, 22B, 22C, 22D.

In the neutron density signal conversion processing mechanism 15, as shown in the flowchart in FIG. 7, the input intermediate region monitor instruction signals 29a, 29b, 29c, 29d, 29e, 29f, 29g, 29h to the bypass signals 23a, 23b, 23c, 23d, 23e, 23f, 23g, 2
Channels bypassed are excluded based on 3h, followed by in-core detector insertion status signals 24a, 24b, 24c, 24d, 24
Channels that are not at all insertion positions are excluded based on e, 24f, 24g, and 24h, and then the intermediate region indicating signals of the remaining channels are multiplied by the correction coefficient according to the range signal. As the correction coefficient C, for example, when converting an instruction signal in the range n into an instruction signal in the range m, the one obtained by the following equation is used.

Cm ← n = 10 ^{(nm) / 2} where range n is a value from 1 to 10 and range m
Is normally set to 1, and calculation is performed. Next, if there is a channel having a large deviation in the intermediate region monitor instruction signal converted to the range 1, it is excluded, and the remaining channels are averaged. The conversion factor for converting the average intermediate region monitor instruction signal obtained in this way into the neutron density signal is the average neutron when the operating range of the neutron source region monitor and the operating range of the intermediate region monitor overlap. It is obtained as a ratio of the density signal 25 and the average intermediate area monitor instruction signal. After the conversion coefficient is calculated, the conversion neutron density signal 30 is obtained by multiplying this coefficient by the average intermediate region monitor instruction signal.

When the core output rises to the upper limit of use of the neutron source area monitor and the in-core detectors 4A, 4B, 4C, 4D are withdrawn, the converted neutron density signal 30 is averaged by the neutron density signal selection mechanism 16. It is selected instead of the neutron density signal 25 and recorded as time series data in the recording device 5 together with the time.
The reactivity calculation device 6 inputs the latest converted neutron density signal sequentially from the time-series data recording device 5, calculates the reactivity of the core in real time in the same manner as when the output of the neutron source region monitor is used, This result is displayed on a display device such as the CRT screen 7.

Next, FIG. 8 shows an embodiment of the present invention utilizing the output of the detector of the neutron source region monitor extracted outside the core when the output of the core further increases and reaches the upper limit of use of the intermediate region monitor. It will be explained based on. In the figure, the same parts as those in FIGS. 1 and 6 are designated by the same reference numerals. Since the configuration is almost the same as that in FIG. 1, the operation will be described below.

The in-core detectors 3A, 3B, 3C, and 3D of the neutron source region monitor are completely withdrawn from the core 2 of the reactor 1 when the core output rises to the upper limit of use of the neutron source region monitor, Although it is installed at a position of 60 cm, the core output is about 10% of the reactor rated output near the upper limit of use of the intermediate region monitor, which is about 10% of the upper limit output of the neutron source region monitor at all insertion positions. ^Since it is equivalent to ⁵ times, it becomes possible to measure the neutron flux that decays with distance from the core.
The detector of the neutron source monitor thus installed outside the core
The output signals from 3A, 3B, 3C, 3D are the amplifiers 8A, 8B, 8C,
After passing through 8D, the logarithmic conversion circuits 9A, 9B, 9C and 9D convert the logarithmic coefficient rate signals 20A, 20B, 20C and 20D into the sampling processing mechanism 10. In the sampling processing mechanism 10, the neutron density signals 22A, 22B, 22 of the original count rate are sampled at a constant cycle according to the sampling cycle signal 21.
C and 22D are output. In the averaging processing mechanism 11, the neutron density signal is input and the bypass signals 23A, 2
3B, 23C, 23D and in-core detector insertion status signals 24A, 24B, 24
C, 24D and the converted neutron density signal 30 from the neutron density conversion processing mechanism 15 are input, processed as shown in the flowchart in FIG. 9, and a corrected neutron density signal 31 is output. That is, the neutron density signals 22A, 22B, 22C, 22D are excluded from the calculation of the channel being bypassed based on the bypass signal, and then the in-reactor detector insertion state signal 24
It is confirmed by A, 24B, 24C, and 24D that the detectors of the remaining channels are all in the full extraction position, and the channels not in the total extraction device are installed in the full extraction position. Then, among the neutron density signals, those with a particularly large deviation are excluded,
The average neutron density signal is calculated. However, since this average neutron density signal is the value obtained by the full extraction device of the detector, it is necessary to correct it to the neutron density signal at the time of full insertion. The attenuation correction coefficient is obtained as the ratio of the converted neutron density signal 30 at the output within the use range of the intermediate region monitor and the above-mentioned uncorrected average neutron density signal, so multiply this coefficient by the uncorrected average neutron density signal. The corrected neutron density signal 31 is obtained by.

When the reactor output is within the operating range of the intermediate region monitor, the converted neutron density signal 30 is selected by the neutron density signal selection mechanism 16, but when the reactor output exceeds the upper limit of the operating range of the intermediate region monitor, the corrected neutron density is corrected. The signal 31 is selected and recorded together with the time signal 26 in the time-series data recording device 5 as time-series data. Reactivity calculator 6
Inputs the latest neutron density signal sequentially from this time-series data recorder 5 and calculates the reactivity in real time.
It is displayed on a display device such as the screen 7.

Since the corrected neutron density signal 31 used in this embodiment can be calculated even at the reactor rated output, as shown in FIG. 10, any range over the output region from the neutron source region at the time of full insertion of the control rods through the intermediate region Neutron density can be measured for neutron flux levels of.

That is, from the state where the control rod is fully inserted into the core until the criticality is reached, the output signals from the in-core detectors 3A, 3B, 3C, 3D of the neutron source region monitor at the full insertion position are used, The average neutron density signal 25 based on this output signal is input to the reactivity calculation device 6. When the reactor power rises and reaches the upper limit of use of the neutron source area monitor, reduced neutrons based on the output signals from the in-reactor detectors 3a, 3b, 3c, 3d, 3e, 3f, 3g, 3h of the intermediate area monitor The density signal 30 is selected by the neutron density signal selection mechanism 16 and used for the reactivity calculation. When the output further increases and reaches the upper limit of use of the intermediate region monitor, the neutron source region monitor can measure at all extraction positions, so the corrected neutron density signal 31 obtained from this replaces the converted neutron density signal 30. Used for reactivity calculation. Since all the obtained neutron density signals correspond to the count rate of the neutron source area monitor, the reactivity can be continuously monitored.

It should be noted that above the upper limit of use of the intermediate output monitor, instead of using the output of the neutron source region monitor at the full extraction position, the output of the output region monitor can be used.

[Effects of the Invention] As is apparent from the above description, the reactivity monitoring method and apparatus of the present invention utilize the output signal from the in-reactor detector to fully insert the control rods into the core. The reactivity can be presented to the operator in real time over a wide range of power levels, from a large negative reactivity at time to the reactivity when the reactor is operating at power. Therefore, the labor of the operator in monitoring the nuclear safety of the nuclear reactor can be significantly reduced, and the control rod can be operated quickly and easily.

Further, in the present invention, since the neutron density signal from the inside of the reactor that is sampled and averaged is recorded as time-series data, a measurement mechanism of the neutron level signal in the reactor,
The mechanism for predicting the neutron density signal and the mechanism for estimating the reactivity can be separated. Therefore, the reactivity monitoring of the reactor can be processed in a distributed and asynchronous manner,
It is possible to freely select the method and system configuration of the process of predicting the neutron density signal and the process of estimating the reactivity. Further, it becomes possible to monitor at a plurality of points, and it becomes easier to perform maintenance work, testing and updating of the reactor reactivity monitoring device.

[Brief description of drawings]

FIG. 1 is a block diagram showing an embodiment of the present invention utilizing the output of a neutron source monitor, and FIGS. 2 and 3 are embodiments of the process of the reactivity calculation performed by the reactivity calculation device of the present invention. FIG. 4 (a) and FIG. 4 (b) are graphs showing a temporal change of the reactivity in the critical proximity at the time of reactor startup obtained by the present invention and a control rod operation example corresponding thereto, FIG. 6 is a graph showing an example of control rod operation at critical proximity by conventional core reactivity monitoring, FIG. 6 is a block diagram showing one embodiment of the present invention using the output of an intermediate region monitor, and FIG. 7 is the present invention. 8 is a flow chart showing an example of conversion processing in the neutron density signal conversion mechanism, FIG. 8 is a block diagram showing one embodiment of the present invention utilizing the output of the neutron source region monitor pulled out of the core, and FIG. 9 is the present invention. In the averaging processing mechanism of Flow chart illustrating an embodiment of a process of obtaining a positive neutron density signal, Figure 10 is a graph showing the relationship between the area monitor to be used in the reactor power and the present invention, the 11
Figure is a block diagram showing an example of a conventional reactivity monitoring apparatus, FIG. 12 is a graph showing the time change after the control rod withdrawal operation of the logarithmic count rate by the neutron source monitor at the time of supercritical,
FIG. 13 is a graph showing the time variation of the logarithmic count rate by the neutron source monitor in the subcritical state after the control rod withdrawal operation was interrupted. 1 ... Reactor 2 ... Core 3 ... In-core detector 4 ... Neutron density measuring device 5 ... Time-series data recording device 6 ... Reactivity calculation device 7 ... CRT screen 8 ... Amplifier 9 ... Logarithmic conversion circuit 10 …… Sampling processing mechanism 11 …… Averaging processing mechanism 12 …… Gain adjuster 13 …… Square averaging circuit 14 …… Phase inverter 15 …… Neutron density signal conversion processing mechanism 16 …… Neutron density signal selection Mechanism 20 …… Logarithmic count rate signal 21 …… Sampling period signal 22 …… Neutron density signal 23 …… Bypass signal 24 …… In-core detector insertion status signal 25 …… Average neutron density signal 26 …… Time signal 27 …… Range signal 28 …… Middle region monitor output signal 29 …… Middle region monitor instruction signal 30 …… Reduced neutron density signal 31 …… Corrected neutron density signal

Claims (4)

[Claims] 1. A neutron density measuring device for sampling and averaging output signals from a plurality of in-reactor detectors for measuring neutron flux levels, and outputting an average neutron density signal,
A time-series data recording device that records the latest neutron density signal from this neutron density measuring device as time-series data, and the latest neutrons by inputting the data group of the neutron density signal from this time-series data recording device. The most probable value of the density signal is calculated by function approximation, the external neutron source value is determined, and the neutron density signal corrected by this external neutron source value and the function approximation is applied to the one-point reactor approximate dynamic characteristic equation to react the core. And a reactivity calculation device that calculates the degree of reaction in real time. 2. The external neutron source value is a core reactivity obtained from an eigenvalue of core calculation based on a diffusion equation when the reactor is in a subcritical steady state, and a neutron density signal corrected by the function approximation to a one-point reactor approximation The reactor reactivity monitoring apparatus according to claim 1, wherein the reactor reactivity monitoring apparatus is obtained so as to match the reactivity obtained by using the characteristic equation. 3. The external neutron source value is a core calculation based on a diffusion equation in advance by changing the reactor water temperature, core average control rod density, core average xenon concentration and core average burnup as variables of core state. From the table of eigenvalues obtained by doing
Eigenvalues corresponding to the current reactor water temperature, core average control rod density, core average xenon concentration and core average burnup are determined by interpolation, and the core reactivity obtained from this and the neutron density corrected by the function approximation are obtained. The reactor reactivity monitor according to claim 1, wherein the reactor reactivity monitor is obtained so as to match the reactivity obtained by using the signal in the one-point reactor approximate dynamic characteristic equation. 4. An averaging process step of averaging output signals from a plurality of in-reactor detectors for measuring a neutron flux level to calculate an average neutron density signal, and the average neutron periodically calculated. A data accumulation step of accumulating the density signal as time series data, a neutron density correction step of calculating the most probable value of the latest neutron density signal from the time series data of the accumulated average neutron density signal by function approximation, and the reactor. In the subcritical steady state, the reactivity of the current core state is calculated from the eigenvalue obtained by the core calculation based on the diffusion equation,
An external neutron source calculation step of calculating the fixed external neutron source value by applying the calculated reactivity and the latest neutron density signal corrected in the neutron density correction step to the one-point reactor approximate dynamic characteristic equation, and this external Apply the fixed external neutron source value calculated in the neutron source calculation step to the one-point reactor approximate dynamic characteristic equation to calculate the reactivity of the core in real time from the latest neutron density signal corrected in the neutron density correction step And a reactivity calculation step for performing the same.