JPH0731270B2 - Online reactor power distribution monitoring system - Google Patents

Description

Description: TECHNICAL FIELD OF THE INVENTION The present invention relates to an online power distribution monitoring device for a nuclear reactor.

[Technical background of the invention and its problems]

Generally, in a nuclear reactor, it is indispensable for safe and efficient operation of the reactor to monitor whether or not the core is performing properly while maintaining its soundness. For this reason, an online power distribution monitoring device is usually installed in the reactor, and it becomes possible to estimate the power distribution inside the reactor, which is one of the main parameters for monitoring the soundness of the reactor at any time. ing.

By the way, conventionally, as a means for estimating the power distribution in the reactor, first, the relationship between the count value of the neutron flux measuring instrument in the core and the output of the fuel assembly adjacent to this is evaluated in advance by offline calculation, The result is stored as a fitting formula in the power distribution monitoring device, then in the online power distribution monitoring, the count value of the in-core neutron flux measuring instrument is input, and the output of the fuel assembly is calculated by the fitting formula. Therefore, the power distribution in the reactor is estimated.

However, in such a conventional means, when the measurement error when counting the neutron flux by the in-core neutron flux measuring device is small enough to be ignored, it is possible to estimate the in-core power distribution with high accuracy, Actually, the measurement error in this case cannot be ignored, and it may not always be possible to accurately estimate the power distribution in the reactor. The biggest reason for this is that the measuring instrument deviation is due to the fact that the conduit guiding the measuring instrument is displaced from the prescribed position due to the thermal effect of the core, or the measuring instrument itself is placed in an eccentric position in the conduit. It is unavoidable due to the operation of the furnace. Due to this, when estimating the power distribution in the reactor by the conventional method, an unavoidable and non-negligible error occurs, which needs to be considered as an operating margin, which hinders efficient operation of the reactor. .

Incidentally, in recent years, in place of the above means, an output distribution calculation device having a built-in physical model relating to the output of each part of the core is provided, so that the in-core power distribution calculated by the same matches the count value of the in-core neutron flux measuring instrument A method to estimate the power distribution in the reactor by fitting a physical model has been considered.
Until now, it does not necessarily have a means for effectively correcting the measurement error contained in the count value of the above-described in-core neutron flux measuring instrument, and merely estimates the power distribution assuming that the measurement value is entirely true, Therefore, the above-mentioned drawbacks have not been eliminated yet.

That is, accurate output distribution estimation is possible only in a special case where the output distribution calculation error can be regarded as being the same in all strings, but usually this is not always the case because it is not so. There was a problem that the output distribution could not be estimated.

[Object of the Invention]

The present invention has been made in view of the above circumstances, in the one equipped with a core power distribution calculation device incorporating a physical model of the core, most of the measurement error included in the count value of the core neutron flux measuring instrument The accurate online power distribution can be estimated by correcting the error, and efficient operation of the reactor can be achieved, and high reliability can be obtained even when the measurement error is biased or there are few measurement points. An object is to provide an online output distribution monitoring device.

[Outline of Invention]

The reactor online power distribution monitoring device according to the present invention is
A core neutron measuring instrument installed in each of a number of strings arranged in the core to measure the neutron flux, and a core current status measuring core current data such as core cooling water flow rate, core pressure, core inlet / outlet temperature, and control rod position. A data measuring instrument, and an output distribution calculating device for calculating the power distribution in the core based on the physical model relating to the count value input from this core current state data measuring instrument and the power density of the core built-in in advance,
Input the output distribution calculated by this output distribution calculator,
Based on the power distribution, the power density at a certain position in the core radial direction is integrated in the core axial direction to obtain an averaged value as the core radial direction component of the calculated value, while being provided in the string corresponding to the core radial direction position. Entered the count value from the in-core neutron measuring instrument, when the core radial direction component of the count value is a value obtained by integrating and averaging each count value in the core axial direction along the string, the inside of the core A value obtained by dividing the count value from the neutron measuring device by the core radial direction component of the count value is extracted as the core axial direction component of the count value, and the core radial direction component of the calculated value and the core axial direction component of the count value An output distribution estimation device that corrects the output distribution calculated by the output distribution calculation device so as to match the product with the evaluation value as the evaluation value is provided.

(Operation) A reactor online power distribution monitoring apparatus according to the present invention is
The reactor power can be estimated with high accuracy using the online power distribution. The power distribution in the reactor core is
It is possible to calculate and calculate the calculated value calculated by the physical model built into the power distribution calculation device based on the current core data, while it is calculated by individually calculating the count value measured by the in-core neutron measuring instrument You can also

Now, assuming that the axial position of the core is K and the radial position is L, the count value (power density) which is the output signal from the in-core neutron measuring instrument at the core measurement position (K, L) is ERM (K, L). )age,
The average value at the core radial position L is represented by ERMR (L).
This ERMR (L) is the count value ER at the core radial position L
It is a value (arithmetic mean count value) obtained by integrating M (K, L) in the core axis direction and averaging it, and this value is called the core radial component of the count value.

In addition, the count value ERM (K, L) at the core radial position L is
Value ERMZ divided by core radial component ERMR (L) of count value
(K, L) is called the core axial component of the count value. In this way, the count value can be obtained by dividing the core radial component ERMR (L) and the core axial component ERMZ (K, L).

On the other hand, the core power distribution calculated based on the physical model built into the power distribution calculator is input to the power distribution estimator and calculated to calculate the core measurement position (K, L).
The calculated value (power density) ERC (K, L) for the count value of the in-core neutron measuring instrument at can be calculated. From this calculated value, the calculated value (power density) at the core radial position L
Calculate the average value (arithmetic mean reactor power) ERMC (L) of the value and calculate the radial component ERMR (L) and core axial component ERMZ of the count value.
By the same processing as (K, L), the calculated value is the core radial component E
It can be calculated separately for RCR (L) and core axial component ERCZ (K, L).

Next, based on the physical model of the power distribution calculation device, the core radial direction component of the calculated value obtained by calculation with the power distribution estimation device
ERCR (L) and core axial component ERCZ (K, L) and core radial component ERMR (L) and core of the count value (measured value) obtained by inputting the count value from the in-core neutron flux measuring instrument The axial component ERMZ (K, L) is compared with the most reliable experimental value of the gamma scan experiment of the fuel after irradiation to confirm the reliability of the calculated value and the counted value.

As a result of this comparison, the calculated value (see FIG. 6) for the radial component of the core has less error variation than the count value (measured value) (see FIG. 5), has good accuracy, and is highly reliable. I understood it. As for the core axial component, it was found that the count value (measured value) (see FIG. 4) had a sufficient measurement accuracy than the calculated value. On the contrary, the core radial component ERMR (L) of the count value lacks reliability of the absolute value of the neutron flux distribution, and the core axial component ERCZ (K,
L) was found to have poor neutron flux distribution shape reliability.

The present inventor has adopted a highly reliable and accurate calculated value for the core radial component, and has adopted a count value (measured value) with sufficiently high measurement accuracy for the core axial component, and outputs ERMR of radial direction component of calculated value by distribution estimation device
(L) and count value (measured value) in the axial direction of the core ERMZ (K,
When the evaluation value is obtained from the product of L), this evaluation value can be very closely approximated to the ideal count value (best count value) from the in-core neutron measuring instrument. Power Distribution of Core (Power Density Distribution)
Is corrected to obtain the output distribution with high accuracy for each of the four fuel assemblies.

That is, when obtaining the power distribution in the reactor core, as shown in FIG. 8, the calculated value is reliable and accurate in the core radial component ERCR (L) (see FIG. 8 (a)). Is adopted, and the unreliable core axial component ERCZ (K, L) (see FIG. 8 (b)) is discarded. In addition, regarding the measured values, the core radial component ERMR (L), which has low reliability (Fig. 8 (c))
reference. ) Is discarded, the measurement accuracy is high and the core axial component ERMZ (K, L) with high reliability is shown in FIG. 8 (d). ) Is adopted. All of them adopt reliable and accurate core radial component ERMR (L) of calculated value and core axial component ERMZ (K, L) of measured value (count value), and output the product of them. Evaluated value ERB (K, L) calculated by distribution estimation measurement (Fig. 8 (e))
reference. ) Is calculated, and the core power distribution from the power distribution calculation device is corrected by this evaluation value, and the core power distribution is calculated with high accuracy for each fuel assembly of four sets.

Example of Invention

An embodiment of an online power distribution monitoring device for a nuclear reactor according to the present invention will be described in detail below.

FIG. 1 shows a schematic system of an online power distribution monitoring device, in which a reactor core current data measuring device 2 is provided in the reactor 1 to measure core cooling water flow rate, core pressure, core inlet / outlet temperature, control rod position, etc. , And those measurement value signals are output to the data sampler 3. A fixed neutron flux measuring instrument 5 (Local Power Range Moniter, hereinafter abbreviated as LPRM) and a movable neutron measuring instrument 6 (Trav) are installed in the core 4.
ersing Incore Probe (hereinafter abbreviated as TIP) is provided via a conduit in a fixed and movable state, respectively. As shown in FIG. 2, a large number of conduits 7 (hereinafter referred to as strings) are arranged in the corner gaps 9 between the four fuel assemblies 8 (hereinafter referred to as bundles), and LP
RM5 are arranged in each core string 7 in the axial direction of the core, that is, a plurality of spaces (for example, 4) are vertically spaced, and the TIP6 is selectively inserted into and removed from each string 7 and slides vertically to scan along the core axial direction. Allows continuous neutron flux distribution to be measured. A control rod 10 is arranged on the side opposite to the string of each of the four fuel assemblies 8.

The neutron flux count value signals obtained by LPRM5 and TIP6 are output to the data sampler 3. The data sampler 3 transmits the count value signals from the LPRM 5 and the TIP 6 to the output distribution estimation device 11. The output signal of the LPRM5 is always generated, but the data sampler 3 can output the output distribution estimating device only when there is a transmission request from the output distribution calculating device 12.
Perform transmission to 11. Further, in-core neutron flux measurement by TIP6 is usually performed once a month, and in this case, the data sampler 3 always transmits the output signal of TIP6 to the output distribution estimation device 11.

The data sampler 3 also inputs the measured value signals of the core cooling water flow rate, the core pressure, the core inlet / outlet temperature, the control rod device, etc., which are output from the core current state data measuring device 2, and transmits these to the power distribution calculation device 12. . Although these output signals are constantly generated, the transmission is performed at the same time as the transmission of the LPRM output signal or the TIP output signal to the output distribution estimation device 11.

The power distribution calculation device 12 is automatically started according to a predetermined cycle, or the input / output device 13 of the reactor operator
It is activated by a request via the or via a signal and a trigger indicating the end of in-core neutron flux measurement by TIP6. And for the data sampler 3, TI
A request for transmission of the output signals of P6 and LPRM5 to the power distribution estimation device 11 is made, core current data from the data sampler 3 is input, and the power distribution in the core is calculated by the built-in physical model. As a physical model used for such a calculation, in order to improve the responsiveness of the entire apparatus, the calculation time is short even if the calculation accuracy is a little poor, and a certain limited group of points of the bundle 8 is extracted. It is desirable to use the so-called group-coarse grid point diffusion model. That is,
When the power distribution calculation in the reactor is performed using this group / rough lattice point diffusion model, the bundle 8 shown in FIG. 2 is represented by a set of spatially many nodes (nodes) 8a, and each node Let the neutron flux at satisfy the following equation (I).

L · φ + B ² · φ = 0 (I) L: a square matrix with the number of rows and columns equal to the total number of nodes, which is a finite difference approximation of the Laplace operator φ: with the entire neutron flux at each node as an element A vector of dimensions equal to the total number of nodes.

B ² : A square diagonal matrix with the number of rows and columns equal to the total number of nodes, with the material buckling of each node as a diagonal element.Furthermore, the reactor core is described in a three-dimensional rectangular coordinate system and The coordinates taken are z, and the coordinates taken in the core radius direction are represented by x and y, respectively, and the Kth node in the z direction, the Ith node in the x direction, and the Jth node in the y direction are subscripts K, I, Let us denote by J, and the neutron flux of the node that is the element of φ is φ (K, I,
J). Then, the power distribution in the reactor is given as a set consisting of the power density at each node, P (K, I, J) = Σ _f (K, I, J) · φ (K, I, J) (II). To be Where P (K, I, J): power density at node (K, I, J), where the unit is the number of nuclear fission per unit time and unit volume Σ _f (K, I, J): node (K, I) , J) Macroscopic Fission Cross Section Based on the above equations (I) and (II), the power distribution calculation device 12 calculates the power distribution in the reactor by a known method. This calculation method is described in, for example, M. Tsuiki et al., Published in pages 724 to 732 of Nucl. Sci. Eng., 64 (published in 1977) in the United States; “Convergence and Acceleration of Void Itertions i
n Boiling Water Reactor Core Calculations ". The power distribution calculation device 12 transmits the calculated in-reactor power distribution to the power distribution estimation device 11.

The output distribution estimation device 11 transmits the data sampler 3.
Output signal of TIP6 and LPRM5, and output distribution calculator 1
Taking in the calculated in-reactor power distribution in step 2, first estimate the TIP count value as follows.

That is, from the in-reactor power distribution (P (K, I, J) in the formula II) transmitted from the power distribution calculation device 2, T
Calculate the IP count.

ERC (K, L) = Sum C (K, I, J) x (K, I, J) (III) ERC (K, L): Core axial node position K, string L
Calculated value for the TIP count value at the measurement position (K, L).

Sum: Sum of all nodes adjacent to the measurement position (K, L) C (K, I, J): Conversion coefficient for converting the output density at the node into a TIP coefficient where C (K, I, J) is calculated Some have been devised, but there are the following methods with high accuracy. That is, first, FO (K, L) = {Sum FINF (K, I, J) × κ (K, I, J)} ÷ {Sum κ (K, I, J)} (IV) FINF (K, I , J): Asymptotic value of thermal neutron flux at the node,
This is obtained by usual fuel assembly nuclear constant calculation κ (K, I, J): The reciprocal of the diffusion distance of thermal neutrons at the node, which is obtained by ordinary fuel assembly nuclear constant calculation FO (K, L ): Measuring position of the gap between fuel assemblies (K, L)
Compute the thermal neutron flux at.

In addition, this formula (IV) is described in “Tsuki Castle“ Study of Nuclear Calculation Methods in Boiling Water Reactor Core Characteristics Analysis ”, Kyoto University Dissertation (19
77)]-This is abbreviated as Reference 1 below-in formula (3.23), (1) the fast neutron flux is spatially uniform, and (2) the diffusion distance of thermal neutrons is Compared with the width Δ, it can be regarded as coth (κ ・ Δ / 2) ~ 1 and is easy to use, taking advantage of the fact that (3) the diffusion coefficient of thermal neutrons is spatially uniform. Can be obtained.

On the other hand, the thermal neutron flux φ '(K, L) at the measurement position (K, L) of the gap between the fuel assemblies can be obtained by ordinary calculation of the nuclear constant of the fuel assembly. It is calculated based on the assumption that the resulting fuel assemblies form an infinite lattice system, and when fuel assemblies with different compositions are adjacent to each other as in the actual core, It cannot be said that the value is correct and correction is necessary.

So here, this correction is given by equation (IV)
Use FO. According to Reference 1, the formula (I
FINF appearing on the right side of (V) is the value of the thermal neutron flux calculated based on the assumption that the fuel assemblies subjected to the calculation of nuclear constants form an infinite lattice system. ) Calculated by) represents the thermal neutron flux at the measurement position (K, L) of the gap under the condition that fuel assemblies with different compositions are adjacent to each other in the actual core. . That is, the thermal neutron flux that was FINF (K, I, J) in the infinite lattice system changes FO (K, L) at the actual measurement position (K, L) in the core. So this change is FO and F
Evaluated by the ratio of INF, φ (K, L) = {FO (K, L) / FINF (K, I, J)} × {P (K, I, J) / PI (K, I, J) )} × φ ′ (K, I, J) (V). However, here PI (K, I, J) is the power density of the node used in the calculation of the fuel assembly nuclear constant,
The correction to take into account that the magnitude of the thermal neutron flux is different because it is different from the actual power density P (K, I, J) in the core is ratio {P (K, I, J) / PI ((K, I,
J)}. Because most of the fission that occurs in the core is caused by thermal neutrons, the magnitude of thermal neutron flux is proportional to the magnitude of power density.

From the above, the thermal neutron flux φ (K, L) at the measurement position (K, L)
, Which can be calculated only from the quantity unique to one of the nodes adjacent to the measuring instrument. Therefore, for the present purpose, equation (V) may be calculated for all the nodes adjacent to the measuring instrument of interest, and the arithmetic mean thereof may be taken. That is, the measurement position (K,
Thermal neutron flux φ (K, L) at L) is φ (K, L) = Sum [{FO (K, L) / FINF (K, I, J)} × {P (K, I, J) / PI (K, I, J)} × φ ′ (K, I, J)] ÷ Sum [1.0]. The count of the neutron flux measuring instrument at the measurement position (K, L) is obtained by multiplying the thermal neutron flux obtained here by the conversion factor H (K, L) peculiar to the measuring instrument. , L) = H (K, L) × Sum [{FO (K, L) / FINF (K, I, J)} × {P (K, I, J) / PI (K, I, J)} × φ '(K, I, J) ÷ Sum [1.0]. From this formula and formula (III), C (K, I, J) = H (K, L) × [{FO (K, L) / FINF (K, I, J)} ×× {1.0 / PI ( It can be seen that K, I, J)} x φ '(K, I, J)] ÷ Sum [1.0] (VI).

From the above, the calculated value ERC for the count value of the neutron flux measuring instrument for all measurement positions in the core can be obtained, but these are obtained independently from the measuring instrument which is the output signal of the in-core neutron flux measuring instrument described below. Is.

Next, the output distribution estimation device 11 extracts the core axial direction (z direction) component from the output signals of the TIP6 and LPRM5 transmitted from the data sampler 3. First, the case of TIP6 will be described.

Now, assuming that the axial position of the core is K and the radial position is L, the output signal (count value) of the core neutron measuring instrument TIP at the core measurement position (K, L) is ERM (K, L). And the average value of the count value (TIP output signal) at the core radial position L is ERMR (L), This ERMR (L) is an arithmetic mean value obtained by integrating and averaging the TIP count values at the core radial direction position L in the core axis direction, and this arithmetic mean value is called the core radius direction component of the count value.

TI corresponding to the power density at the core radial position L
P count value ERM (K, L)
If the value divided by (L) is ERMZ (K, L), it is expressed as ERMZ (K, L) = ERM (K, L) / ERMR (L), and this ERMZ (K, L) is calculated as Since the information about the numerical value in the radial direction component ERMR (L) is lost, this is called the core axial direction component of the count value.

If the equation expressing the core axial component ERMZ (K, L) of this count value is modified, it is expressed as ERM (K, L) = ERMR (L) · ERMZ (K, L), and the TIP output signal (output density ) Is a count value ERM
It can be seen that (K, L) can be divided into a core radial component ERMR (L) and a core axial component FRMZ (K, L) of the count value.
In this formula, ERMR (L) is the arithmetic average of the output signals of the string L of interest in all axial directions.
It therefore represents the average magnitude of the output signal in the string. Since the strings are installed almost uniformly in the core radial direction, ERMR can also be interpreted as a component representing the core radial distribution of the TIP count value. On the other hand, ERMZ (K, L) is the original output signal ERM (K, L) divided by the average value ERMR (L) of the string. It loses information about the component, and is a quantity that represents only the axial component of the TIP count value in that string. Here, it is said that the ERMZ (K, L) is calculated by this operation to extract the core axis direction component of the TIP count value, which is the TIP output signal.

Next, the power distribution estimation device 11 extracts the core radial component from the calculated value ERC (K, L) for the TIP count value calculated by the equation (III) by the following means. That is, ERCZ (K, L) = ERC (K, L) / ERCR (L) From these equations, the calculated value for the TIP count value is calculated as ERC (K, L) = ERCR (L) ・ ERCZ (K, L).
You can see that it can be divided into (L) and ERCZ (K, L).
Here, ERCR (L) is the same as the above-mentioned ERMR can be interpreted as a component that expresses the core radial distribution of TIP count values.
It can be interpreted as a component that represents the core radial distribution of the calculated value with respect to the count value. ERCZ (K, L) can be interpreted as a core axial component. Here, the ERCR (L) is calculated by this operation to extract the core radial component of the calculated value for the TIP count value.

Next, the output distribution estimation device 11 uses the following equation to calculate the TIP count value ERB
Estimate (K, L). That is, ERB (K, L) = ERCR (L) · ERMZ (K, L) (VII). As known from this formula, the best TIP count ERN
(K, L) takes the core radial component ERCR (L) from the calculated value for the TIP count value, while the output signal of TIP, TIP
It is calculated by taking the core axial component ERMZ (K, L) from the count value and multiplying them.

The best TIP count here means as close as possible to the ideal TIP output signal that would be obtained if there were no measurement errors involved in the measurement.

Therefore, by using this best TIP count value as an "evaluation value" and correcting the output distribution with this, a highly accurate output distribution can be obtained.

Note that, hereinafter, this “evaluation value” will be referred to as the “best TIP count value” and described.

In estimating the above-mentioned best TIP count value, the core radial component is discarded from the output signal of TIP, and the reason for using only this core axial component is that the measurement error due to TIP6 is mainly the core of TIP6 or its conduit. It is due to the deviation in the radial direction. This will be described with reference to FIG. That is, when measuring the neutron flux by TIP6, the TIP must be guided so as to pass through the accurate center position 14 of the gap 9 between the fuel assemblies 8. However, the conduit 7 of the TIP6 changes depending on the operation of the reactor. Enclosed / deformed, TIP6
It is practically difficult to pass the TIP 6 through the central position 14 due to technical difficulties in positioning the TIP 6 in its own conduit, and there is inevitably a TIP 6 guided by the position 15 deviated from the central position. . On the other hand, the current TIP measures the neutron flux in the core due to the occurrence of uranium-235 fission, so its sensitivity is the maximum for thermal neutron flux and almost no sensitivity for neutron flux in other energy regions. Do not have. However, the thermal neutron flux has an apex at the center of the gap between the fuel assemblies, and exhibits a steep mountain-shaped distribution 16 that rapidly decreases with distance from the apex. This is because neutrons are slowed down and heat-exchanged better than in the fuel assemblies due to light water existing in the gaps between the fuel assemblies. From the above, if the TIP 6 does not pass through the exact center 14 of the inter-fuel-assembly gap 9 but passes through the position 15 deviated from the center, an error 17 occurs in the count value. However, the above-mentioned mountain-shaped distribution of thermal neutron flux 16
Is uniform in the axial direction of the core, this error has a uniform magnitude in the axial direction of the core, and the average magnitude of the output signal of this string 7 is ERMR (L). Is changed from the correct value, but TI in string 7
The core axial component ERMZ (K, L) of the P count value is almost the same as the correct one.

This was confirmed experimentally. That is, in FIG. 4, the z coordinate, that is, the coordinate taken along the axial direction of the core is plotted on the horizontal axis, and the La (lanthanum) 140 concentration at each node of the fuel assembly is plotted on the vertical axis to obtain the true value. Relative to the axial distribution, the experimental value a from the gamma scan experiment after irradiation, which has a high measurement accuracy of about 2%, and the equivalent La140 concentration b converted from the TIP output signal. It is a comparison. As shown in the figure, the agreement between the two is good. On the other hand, in Fig. 5, the horizontal axis is the distance from the center of the core to the fuel assembly, and the vertical axis is the average La140 in each fuel assembly.
For density error, the equivalent L converted from the TIP output signal to the experimental value from the gamma scan experiment after irradiation
The error c of the density of a140 is shown. As can be seen, the error shown here cannot necessarily be ignored. Here, the experimental results shown in FIG. 4 confirm that the core axial component ERMZ (K, L) of the TIP count values are almost the same, and the experimental results shown in FIG. This proves that the core radial component ERMR (L) has an error that cannot be ignored.

For this reason, in estimating the best TIP count value, the core radial component is discarded from the output signal of the TIP, and only the core axial component is used.

On the other hand, in estimating the TIP count value, the core radial component ERCR (L) is taken from the calculated value for the TIP count value, and the core axial component ERCR (K, L) is discarded. It has been experimentally confirmed that the power distribution calculated by the physical model built in the device 2 can accurately calculate the core radial component even if the calculation accuracy of the core axial component is insufficient. Is. That is, FIG. 6 shows the distance from the center of the core on the horizontal axis to the fuel assembly,
The vertical axis represents the error of the average La140 concentration in each fuel assembly, and the equivalent La140 concentration error converted from the output distribution calculated by the physical model to the experimental value of the gamma scan experiment after irradiation. Is. These errors, as seen in the figure, are shown in FIG.
About 1/2 of the error of La140 concentration converted from TIP count value
It is within the following range and can be ignored. This means that the core radial component of the power distribution calculated value by the physical model has good accuracy, and at the same time, the core radial component ERCR (of the calculated value for the TIP count value calculated by the linear transformation (III) for the power distribution) This proves that L) has good accuracy.

For the above reasons, when estimating the best TIP count value, the core radial component ERCR (L) is calculated from the calculated value for the TIP count value.
, And the core axial component ERCZ (K, L) is discarded.

That is, when obtaining the power distribution in the reactor core, as shown in FIG. 8, the calculated value is reliable and accurate in the core radial component ERCR (L) (see FIG. 8 (a)). Is adopted, and the unreliable core axial component ERCZ (K, L) (see FIG. 8 (b)) is discarded. In addition, regarding the measured values, the core radial component ERMR (L), which has low reliability (Fig. 8 (c))
reference. ) Is discarded and the core axial component ERMZ (K, L) (see FIG. 8 (d)) with high measurement accuracy and high reliability is adopted. All of them are reliable and accurate, and adopt the core radial component ERCR (L) of the calculated value and the core axial component ERMZ (K, L) of the measured value (count value) and output the product of them. Evaluated value ERB (K, L) calculated by distribution estimation measurement (Fig. 8 (e))
reference. ) Is calculated, and the core power distribution from the power distribution calculation device is corrected by this evaluation value, and the core power distribution is calculated with high accuracy for each fuel assembly of four sets.

Next, the power distribution estimation device 11 corrects the power distribution calculated value so that the power density of the node matches the best TIP count value estimated above. That is, G (K, I, J) = ERB (K, L) / ERC (K, L) (VIII), PC (K, I, J) = G (K, I, J) P (K , I, J) (IX) Here, the node (K, I, J) refers to everything adjacent to the measurement position (K, L). PC (K, I, J) calculated by formula (IX)
By substituting P (K, I, J) in equation (III), the definition of the correction factor G (K, I, J) and equation (III) clearly reveals that ER
Obtain the best TIP count ERB (K, L) as C (K, L). In this sense, PC (K, I, J) is said to fit the best TIP count.

Finally, the output distribution estimation device 11 uses the PC (K, I,
J) is corrected so that its radial component matches that of the output distribution calculated value by the physical model. That is, PZ (K, I, J) = P (K, I, J) / PR (I, J) to determine P (K, I, J) as core axial component PZ (K, I, J) and core radial direction Divide into components PR (I, J). Considering that the accuracy of the core radial component of the power distribution calculated by the physical model is good as described above, the radial component of PC (K, I, J) is P
Correct it with the following formula so that it matches R (I, J).

PCR (I, J) = Sum PC (K, I, J) /Sum1.0 N (I, J) = PR (I, J) / PCR (I, J) PB (K, I, J) = N (I, J) · PC (K, I, J) (X) Here, PB (K, I, J) calculated by the equation (X) is finally obtained by the output distribution estimation device 11 of the present invention. And this
It is called the best output distribution in the sense that it is estimated by extracting only accurate components from the TIP output signal and the calculated values of the physical model.

The calculation method of the best output distribution described above is one embodiment of the output distribution estimation device in the present invention, and the best output distribution PB
The calculated value for the in-core neutron flux meter count value calculated based on (K, I, J) is equal to the best TIP count value,
The core axial component of the best TIP count value is ERMZ (K, L), which is equal to the core axial component of the count value of the in-core neutron flux measuring instrument. On the other hand, the best output distribution PB (K, I,
The core radial component of J) is designed to match the core radial component of the calculated power distribution by the physical model. Therefore, the best power distribution PB (K, I, J) here is the power distribution P (K, I, J) calculated by the power distribution calculation device that calculates the power distribution in the core based on the physical model. Calculated based on the distribution, the core axial direction component of the calculated value for the count value of the in-core neutron flux measuring device, so as to match the core axial direction component of the count value of the in-core neutron flux measuring device, and also The radial component of the output distribution is obtained by correcting the radial distribution of the output distribution itself so as to match the radial component.

It has been experimentally confirmed that the best power distribution estimated above has better accuracy than the power distribution estimated assuming the TIP output signal to be correct. FIG. 7 shows the z-coordinate, that is, the coordinate taken along the axial direction of the core on the horizontal axis, and the concentration of La140 at each node of the fuel assembly on the vertical axis, and the experiment by the gamma scan experiment after irradiation. The value e,
FIG. 6 is a comparison of the equivalent La140 concentration f converted from the TIP output signal and the equivalent La140 concentration g converted from the output distribution calculated by the device according to the present invention. As can be seen in the figure, the equivalent La140 concentration f converted from the TIP output signal has a distribution shape along the core axis direction that agrees well with the gamma scan experimental value e, but overall The value is too small, confirming the above. On the other hand, the power distribution calculated by the apparatus according to the present invention or the equivalent concentration g of La140 to be converted is in good agreement with the gamma scan experimental value e in both the core axial distribution shape and the overall size. , Supporting the effect of the device according to the invention.

According to the above embodiment, the product of the core radial component based on the calculated value of the power distribution calculation device, and the core axial component using the value obtained by dividing the count value of the in-core neutron flux measuring instrument by its average value. Since the evaluation value is used, the error originally included in the calculated value and the count value is almost completely eliminated. That is, the radial component, which is a highly accurate calculated value obtained by dividing the integrated value of the output distribution by the axial distance, and the highly accurate axial component, which is obtained by dividing each count value by its average value. Since the evaluation values obtained by once decomposing and then multiplying these are used, even when the error is biased in one of + or-directions, or when the number of measurement points is small and the number of data is small. Even if there is, it is possible to perform highly accurate correction according to the actual output, and it is possible to estimate the output distribution with higher reliability than before.

In addition, in the said Example, although the case where the in-core neutron flux measuring device was TIP6 was described, it can implement even when it is LPRM5. However, in that case, only a few measurement points can be obtained in the axial direction of the core, and the output signals at the measurement points K corresponding to all nodes in the axial direction cannot be used as in the case of TIP6. That is, the correction coefficient G (K, I, J) calculated by the equation (VIII) cannot be obtained for all nodes K in the core axis direction, and can be obtained only for K at a few points at most. Therefore, in such a case, G
Correction factor G for nodes for which (K, I, J) cannot be obtained
Estimate (K, I, J). That is, (A) For nodes above the LPRM located at the top in the core axis direction and for which no correction coefficient is required, the correction coefficients found for the nodes adjacent to the LPRM located at the top in the core axis Be G (K, I, J).

(B) For nodes that are below the LPRM at the bottom in the core axial direction and for which no correction coefficient is required, use the correction coefficient obtained for the node adjacent to the LPRM at the bottom in the core axial direction as G (K, I, J).

(C) For a node other than the above and for which a correction coefficient is not obtained, the correction coefficient obtained for the node above the node and adjacent to the LPRM closest to the node, and Let G (K, I, J) be a value determined by linear interpolation with a correction coefficient obtained for a node below the node and adjacent to the LPRM closest to the node.

After calculating the correction coefficient G (K, I, J) for all nodes by such a method, the output distribution PB (K, I, J) is estimated by the same method as the above equation (IX) and thereafter. To do.

In the embodiment described above, an example in which the power distribution calculation device 12 is used for calculation of the power distribution in the furnace has been described, but in the power distribution calculation in the furnace, the present applicant has already filed Japanese Patent Application No. 55-72885.
Using the reactor power distribution predictor presented in
If the core power distribution at the time of monitoring is calculated, the calculation accuracy of the power distribution calculation value P (K, I, J) is greatly improved, and therefore the calculation value ERC (K, L) for the TIP or LPRM count is calculated. ) Calculation accuracy is greatly improved. As a result, the above correction coefficient G (K, I, J) becomes a value extremely close to 1.0, and the above (A)
The estimation accuracy of the correction coefficient estimated by the methods shown in (B) and (C) is greatly improved, and the accuracy of the finally obtained output distribution can be improved.

〔The invention's effect〕

As described above, according to the reactor power distribution monitoring apparatus of the present invention, by using the evaluation value composed of the value of the core axial direction component of the measured value and the core radial direction component of the calculated value, depending on the string It is possible to estimate the power distribution with high accuracy at all times because the correction coefficient can be calculated, and most of the measurement errors included in the count value of the in-core neutron flux measurement device are corrected to estimate the high-precision online power distribution. As a result, the excellent effect of achieving efficient operation of the nuclear reactor can be achieved.

In particular, according to the present invention, the core radial direction component of the calculated value obtained by processing the output distribution of the core from the power distribution calculation device, and the count value obtained by calculating the count value of the neutron measurement value in the core Since the product of the (measured value) and the core axial component is taken as the evaluation value, the core radial direction component of the calculated value and the core axial direction component of the count value with high reliability and high accuracy are respectively adopted for this evaluation value. The error originally included in the calculated value and the counted value in measuring the furnace power distribution can be almost completely removed. That is, the power density at a core radial position of the reactor can be divided into a core radial direction component and a core axial direction component with both the calculated value and the count value, but in the present invention, the reliability of the core radial direction component in the calculated value is Based on the new knowledge that the count value (measured value) of the core component in the axial direction is high and the measurement accuracy is excellent, both the core component of the calculated radial component and the calculated value in the core axial direction are highly accurate and reliable. Since the directional component and is picked up and these are multiplied to obtain the evaluation value, this evaluation value is the count value (the best count value from the ideal in-core neutron measuring instrument assuming that there is no measurement error). ) Is very close to. The output distribution from the output distribution calculator was corrected with this evaluation value using the output distribution estimator, so even if the measurement error is biased in either + or-direction, or if there are few measurement points and the number of data is small. Even if it is, it is possible to make a highly accurate correction to the actual output, and the reliability is higher than that of the conventional online power distribution monitoring device, and the accuracy is high for each four-body fuel assembly. The output distribution is obtained.

[Brief description of drawings]

FIG. 1 is a block diagram showing an embodiment of an online power distribution monitoring device for a nuclear reactor according to the present invention, and FIG. 2 is an external view showing a part of the core for showing the arrangement of neutron flux measuring instruments in the core. Figures and 3 are conceptual diagrams showing how count errors occur during measurement due to deviation of the neutron flux measuring instrument in the core, and Figure 4 shows sufficient measurement accuracy of the axial component of the TIP count value. 5 is a graph showing that the core radial component of the TIP count value has a measurement error that cannot be ignored, and FIG. 6 is a graph showing that the core radial component of the power distribution calculated by the physical model is sufficiently calculated. 7th graph showing accuracy
FIG. 8 is a graph showing the effect of the reactor online power distribution monitoring apparatus of the present invention, and FIGS. 8A to 8E are calculated based on a physical model by the reactor online power distribution monitoring apparatus of the present invention. It is a figure which shows the procedure which calculates | requires a core radial direction component of a calculated value and a core axial direction component of a measured value which are both highly reliable from the calculated value and the measured value actually measured by the neutron measuring device. 1 ... Reactor, 2 ... Core current data measuring instrument, 3 ... Data sampler, 4 ... Core, 5 ... LPRM, 6 ... TI
P, 7 ... String, 8 ... Fuel assembly, 9 ... Fuel assembly gap, 10 ... Control rod, 11 ... Output distribution estimation device, 12 ... Output distribution calculation device, 13 ... Input / output Equipment, 14
...... Center position of the gap between the fuel assemblies, 15 …… Position deviated from the center position of the gap between the fuel assemblies, 16 …… Thermal neutron flux distribution in the gap between the fuel assemblies, 17 …… Neutron flux measuring instrument Error caused by the deviation from the center position of the inter-fuel-aggregate gap of a, a ...
b …… The core axial component of the TIP count was converted to the equivalent La140 concentration core axial component, c …… The core radial component of the TIP count was converted to the equivalent La140 concentration core radial component However, it was obtained by the experiment after irradiation of the fuel assembly.
Error of La140 concentration with respect to the radial component of the core, d ... Although the radial component of the calculated power distribution by the physical model was converted into the equivalent La140 concentration of the radial component of the core, it was determined by the post-irradiation experiment of the fuel assembly. Error of the obtained La140 concentration with respect to the radial component of the core, e ... La140 concentration obtained by the experiment after irradiation of the fuel assembly, f ... Equivalent to the power distribution obtained by the conventional online power distribution monitor Na
La140 concentration, g ... La140 concentration equivalent to the power distribution obtained by the online power distribution monitoring device of the present invention.

Claims (1)

[Claims] 1. An in-core neutron measuring instrument installed in each of a number of strings arranged in the core to measure neutron flux, and the current state of the core such as core cooling water flow rate, core pressure, core inlet / outlet temperature, control rod position, etc. A core current data measuring device for measuring data, and a power distribution calculation device for calculating the power distribution in the core based on the count value input from this core current data measuring device and the built-in physical model regarding the power density of each part of the core , Input the power distribution calculated by this power distribution calculation device, and based on this power distribution, the power density at a certain core radial position is integrated in the core axial direction and averaged, and the calculated value is the core radial component of the calculated value. While obtaining as, while inputting the count value from the in-core neutron measuring instrument provided in the string corresponding to the core radial direction position, the axial direction of the core along the string When the value obtained by integrating and averaging each count value at the position is the core radial component of the count value, the value obtained by dividing the count value from the in-core neutron measuring device by the core radial component of the count value is calculated. Extracted as a numerical value in the axial direction component of the core, the product of the core radial direction component of the calculated value and the core axial direction component of the count value as the evaluation value, calculated by the power distribution calculation device to match the evaluation value. And an output distribution estimating device that corrects the output distribution, and an online output distribution monitoring device for a nuclear reactor.