JPS6232392A - On-line output distribution monitor device for nuclear reactor - Google Patents

Abstract

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

Description

DETAILED DESCRIPTION OF THE INVENTION [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] In general, in a nuclear reactor, while the core maintains its integrity,
Monitoring whether the required performance is being achieved is essential for the safety of nuclear reactors and their efficient operation. For this reason, an online power distribution monitoring device is usually installed in a nuclear reactor, and this monitoring device is one of the main parameters for monitoring the health of the reactor at any time.
It is configured to be able to monitor the power distribution within the reactor.

By the way, the method for calculating the in-furnace power distribution in the conventional online power distribution monitoring system a has been performed in the following manner. That is, first, the relationship between the measured value of the in-core neutron flux measuring device and the output of the adjacent fuel assembly is evaluated in advance by off-line calculation, and the result is built into the power distribution monitoring device as a fitting formula. put.

Next, the measured values of the in-core neutron flux measuring device were input online, and the output of the fuel assembly was calculated using the above-mentioned fitting formula to estimate the in-core power distribution online. However, the conventional in-core power distribution estimation method described above cannot estimate the in-core power distribution with good accuracy when the measurement error when measuring neutron flux by the in-core neutron flux measuring device is so small that it is ignored. I can do it. However, in reality, this measurement error of neutron flux cannot be ignored. This is because deviations of the conduit guiding the measuring instrument or the measuring instrument itself inevitably occur due to reactor operation, and due to this, when using the conventional in-reactor power distribution estimation method,
Since unavoidable errors or errors that cannot be ignored occur, it is necessary to consider these errors as an operating margin.
This hindered the efficient operation of nuclear reactors.

In addition, in recent years, as in the conventional in-core power output t5 estimation method described above, fittings have been developed that have a built-in relationship between the measured value of the in-core neutron flux measuring device and the output of the adjacent fuel assembly. Instead of evaluating using a formula, the online power distribution monitoring device is equipped with a power distribution calculation device that incorporates a physical model of the reactor core, and the in-core power fraction i calculated by this calculation device is calculated by the in-core neutron flux measurement device. A method has been proposed for estimating the in-core power distribution by adapting a physical model to match the value J. However, when using this in-furnace power distribution estimation method to estimate the in-furnace power distribution assuming that the measured values are completely true, it is not possible to accurately estimate the in-furnace power distribution as described above. There was a problem.

[Objective of the Invention 1] The present invention has been made in view of the above circumstances, and its purpose is to obtain the best possible reactor core using a gamma-ray flux measuring instrument whose measurement accuracy hardly changes even with variations in the core radial direction. An object of the present invention is to provide an online power distribution monitoring device for a nuclear reactor that can output the power distribution within the reactor.

"Summary of the Invention" In order to achieve the above object, the present invention provides a core current data measuring device and a power distribution within the reactor core based on a pre-built-in physical model by inputting the measured values from the core current data measuring device. an in-core gamma-ray flux measuring device; and an in-core gamma-ray flux measuring device that inputs the measured values of the in-core gamma-ray flux measuring device, extracts only the core axial component thereof, and calculates the and a best power distribution estimating device that estimates the best power distribution by correcting the core axial component obtained by inputting the power distribution by the core axial component of the measured value of the in-core gamma-ray flux measuring device. This invention relates to an online power distribution monitoring device for a furnace.

[Embodiments of the invention] An embodiment of the present invention will be described with reference to the drawings.

FIG. 1 is a block configuration diagram of an embodiment of the present invention. As shown in the figure, a reactor core 2 is housed in a nuclear reactor 1, and a fixed neutron measuring instrument 3 is housed in this reactor core 2. A variable gamma ray flux measuring device 4 is also provided. In addition, the reactor 1 contains core current data, such as core cooling water flow rate, core pressure,
Measuring devices 5 (representatively indicated by the symbol @5 below) are provided to measure the temperature at the reactor inlet and outlet, the position of the 411 control rod, etc., and the output signals measured by these measuring devices 3 to 5 are used to indicate when the reactor is on. , are input to the line output distribution monitoring device 6. This monitoring device 6 is composed of a data sensor 7, an output distribution measuring device 8, a best output distribution estimating device 9, and an input/output device 10.

By the way, as shown in FIG. 2, the in-core gamma-ray flux measuring device 4 usually has a conduit 13 (hereinafter referred to as a string) installed in a corner gap 12 constituting a fuel assembly 11, and a movable type inside the string 13. A gamma ray flux measuring device 4 is inserted and scans along the core axis direction to measure the continuous gamma ray flux distribution along the core axis direction. Approximately four fixed neutron flux measuring instruments 3 are also installed inside this ring 13. Further, the control rod 14 is inserted into the gap in which the string 13 is installed on the opposite side of the fuel assembly.

Therefore, the data sampler 7 includes a fixed neutron flux measuring device 3 (hereinafter referred to as LPPM) and a movable γ-ray flux measuring device 4 (hereinafter referred to as TIP) installed in the core 2 of the nuclear reactor 1.
and transmits these output signals to the best output distribution estimation device 9. Here, the output signal of the LPPM is always generated, but the data sampler 7 has an output distribution i.
t The transmission is performed only when there is a transmission request from the R device ε3. Furthermore, in-core gamma ray flux measurement using TIP is normally performed at a gradient of about once a month, and in this case, data sampler 7 always transmits the output signal of TIP to best power distribution estimating device 9.

The data sampler 7 inputs output signals from the measuring device 5 that measures core current data, core cooling water flow rate, core pressure, core inlet/outlet temperature, control rod position, etc., and converts these signals into an output distribution calculation device 8. Transmission to J. by the way,
Although the output signals from the measuring device 5 are constantly generated, the transmission of these output signals is performed simultaneously with the transmission of the L P P M output signal or the TIP output signal to the best power distribution estimation device 9 . The power distribution meter p device 8 is activated automatically according to a predetermined period, or upon a request from a reactor operator via the input/output device 10, or measures the neutron flux in the core using the TIP. The output signal of the TIP to the data sampler 7 is automatically activated by using a signal indicating the end of the TIP as -.
A request is made to transmit the output signal of the PPM to the best power distribution estimating device 9, the core current data is input from the data limbler 7, and the in-core power distribution is calculated using the built-in physical model. As a physical model used for such calculations, for example, a so-called 1-group coarse lattice point diffusion model is used, which can shorten the calculation time even if the measurement accuracy is somewhat poor in order to improve the responsiveness of the entire device. is desirable. When calculating the power distribution in the reactor using such a single group/coarse grid point diffusion model, the bundle 11 shown in FIG. It is assumed that the neutron flux at each node satisfies the following equation (I).

L・ΦtoB・Φ=O...(I) Here, a square matrix Φ with the number of rows and columns equal to the total number of nodes, which is a finite-difference approximation of the L nira plus operator: neutron flux at each node Vector B2 with dimensions equal to the total number of nodes, with the whole as an element: material buckling of each node as a diagonal element,
A square diagonal matrix with the same number of rows and columns as the total number of nodes is used to describe the reactor core in a three-dimensional orthogonal coordinate system, with the core axis direction represented by seven axes, and the core radial direction represented by the X and y axes. ,1
The subscript is the node that exists in the direction, the first node in the x direction, and the third node in the y direction.

Let J denote the neutron flux at a node that is an element of Φ.
Let us write it as (K, l, J). Then, the in-furnace power distribution is the power density P (K) at each node.

1, J) is given by the set consisting of the following equation ([).

P(K, 1.J) = Σ(K, I, J) ・Φ (K, 1.J)
... (It) Here, P (K, 1.J): Power density at the node (K, 1.J), where the unit is the number of nuclear fissions per unit time/unit volume (K, I, J) : Macroscopic fission cross section at node (K, r, J).

Based on the above equation (I) and equation (n), the power distribution calculation device 8 calculates the in-furnace power distribution using a well-known method. This calculation method can be used, for example, in ``Invergence and Acceleration of Void IT Telecommunication Inboiling Water Reactor'' by Ryōgai Tsukishi.
Core Power Recycling”Nuclear Science
Engineer 64, 724-732 (1977) [
M, Tsuiki etal: “C0nVer<Ie
nCe and ACCeleratiOnOfV
oid I terations in Boi
lina Water Reactor Core
Calculations u, NLICI.

Sci, Ena, 64, 724-732 (1
977) ] J. The power distribution calculation device 8 transmits the in-core power distribution calculated in this way to the best power distribution estimation device 9.

The best power distribution estimation device 9 takes in the TIP and LPRM output signals transmitted by the data sampler 7 and the in-core power distribution determined by the power distribution calculation device 8, and first estimates the best TIP measurement value as follows. do. That is, first, from the in-furnace power distribution transmitted from the power distribution calculation device 8, the 31r1 value for the corresponding TIP measurement value is calculated using the following equation (I[[).

ERC (K, L) -8umP (K, 1.J) x C (K, l,, J>-・
・・・・・・・・・(I[I) Here, ERC(K,L): Node position K in core axial direction, this is abbreviated as measurement position δ(K,L)) at T [Pillll constant value Calculated value for 3 uLI: Sum C(K, I, J) for all nodes adjacent to measurement position (K, L): 1. By the way, the above meter DIG (K, 1.J) can be strictly determined as follows, for example: be able to. However, the number of fission reactions or absorption reactions of each nuclide in each region of each fuel rod or chitunnel box, etc., obtained from the nuclear constant meter of a fuel assembly in an infinite lattice system, which is normally performed, Determine the energy dependence of the gamma rays generated during unit nuclear fission, absorption, and reaction. Using this γ-ray source, a system with a γ-ray flux detector placed at the center of an infinite lattice system of single-fuel assemblies is constructed.
Perform ray transport calculations, sift out the measured values when the gamma ray flux measuring instrument is surrounded by a single type of fuel assembly, and fit the results with burnup and void fraction as constants for that fuel assembly type. I will remember it. On the other hand, the calculated value for the measured value of the gamma ray flux measuring device for the rated output of each node calculated using the burnup and void ratio of each node and the above-mentioned fitting coefficient is calculated using the string obtained by the output distribution meter n device. By weighting and averaging the outputs of the surrounding fuel assemblies, a calculated value for the measurement value of the gamma ray flux measuring device located at the center of the fuel assemblies surrounding the string can be obtained.

Here, the reason why the output of the fuel surrounding the string is weighted is that the level of the γ-ray source rises and falls linearly depending on the output of each fuel assembly, and that each γ-ray source
This is due to the characteristic of gamma rays that they are independent of the ray flux. Furthermore, by using the results of gamma-ray transport calculations for an infinite lattice of many single fuel assemblies, it is possible to determine the gadolinium concentration and position of the gamma-ray source and gadolinia rods in each fuel rod or region without performing gamma-ray transport calculations. Since the gamma-ray flux detector measurement value can be easily calculated from parameters such as the average enrichment and void fraction of the assembly, it is usually easy to calculate the gamma-ray transport calculation for an infinite lattice system of a single fuel assembly. There will be no need for it.

As explained above, the calculated values ERC for the measured values of the gamma-ray flux measuring device for all measurement positions in the core are obtained, but these are the measured values that are the output signals of the in-core gamma-ray flux measuring device, which will be explained below. can be obtained completely independently.

Next, the best power distribution estimating device 9 extracts components in the core axis direction (seven directions) from the TIP or LPPM output signal transmitted from the data sampler 7. First, the case of VIP will be explained.

ERM the TIP output signal at the measurement position (K, L)
(K, L), the extraction of the core axial component is
It is obtained by executing the following equations NV) and (V).

ERMZ (K, L) = ERM (K, L)/ERMR (L) ・ (IV)
ERMR(L) ......(V) Here, 3 ul: The following equation (Vl) is obtained from the above equations (rV) and (V) for the sum of the total core axial orientation iK. .

ERM (K, L) = ERMR (L)・ERMZ (K-, L)...(V
l) Looking at this formula (Vl), it can be seen that the output signal of TIP can be divided into two components ERMR and ERMZ. However, in this equation (Vl), ERMR(L) is the arithmetic average of the output signals at 43 axial measurement positions of the string L of interest, so it is the average of the output signals at that sling L. It represents the size.

Furthermore, since the strings are installed almost uniformly in the radial direction of the core, the above ERMR(L) can be interpreted as a component representing the distribution of TIP measurement values in the radial direction of the core. On the other hand, 1yo RMZ (K, L) is the original output signal ER
M (K.

) divided by the average value ERMR(L) of that string, it loses information about the core radial component of the TIP measurement value in the above sense, and the TIP measurement for that string This value represents only the axial component of the core. Here, this operation will create the ERMZ
By outputting (K, L) n, the core axis direction component of the TIP measurement value, which is the TIP output signal, is extracted.

Next, the best output distribution estimating device 9 calculates the calculated value ERC (K-, L
), the core radial direction component can be extracted by executing the following equations (Vl) and (■). In other words, ERCZ(K,L) =ERC(K,L)/ERCR(L)...(Vl)
FRCR(L) = Sun+E RC(K, l-) /Sum[1, O
] - to K... (■) From these formulas (VI) and (■), the following formula (IX) is obtained.

ERC(K,L,) =ERCR(1)・ERCZ(K,l-)...(IX
) Looking at this equation (IX), we can see that the calculated value for the IP measurement value is divided into two components ERCR (L) and ERCZ (K.

It can be seen that it can be divided into L). Here ERCR(L)
can be interpreted as the component representing the core radial distribution of the TIP3111 constant value.
It can be interpreted as a component representing the core radial distribution of calculated values with respect to IP measured values.

Therefore, here, the core radial component of the calculated value for the T[P311 constant value is extracted by calculating ERCR (shi) using the above equation (■).

Next, the best output distribution estimating device 9 estimates the best TIP measurement value ERB(K,L) using the following equation (X). That is, FRB (K, L) = ERCR (L)・ERMZ (K, L)...(X)
As can be seen from this equation (X), the best TIP measurement value ERB
(K, L) takes the core radial component ERCR (L) from the calculated value for the TIP measurement value, and the core axial component ERCR (L) from the TIP measurement value, which is the output signal of the TIP.
It is calculated by taking MZ (K, L) and multiplying both components. Here, the best TIP measurement value is meant to be as close as possible to the ideal TIP output signal that would be obtained assuming that there were no measurement errors involved in the measurement.

By the way, in estimating the above best TIP measurement value, TI
The reason why only the core radial component is used from the output signal of P is that almost no error in measurement by the TIP occurs due to the deviation of the TIP or its conduit at a fixed set point on each side of the core radial direction. This will be explained using FIG. 3. That is, when measuring gamma ray flux using a TIP, the TIP must be guided so as to pass through the exact center position 15 of the gap 12 between the fuel assemblies 11. It is difficult in practice due to variations and deformation caused by operation, and the technical difficulty of positioning the TIP itself within the conduit, and it is difficult to set the position 16 that is off center to a straight line or a position curved with respect to the center. There will inevitably be a TIP guided by 17. However, since the gamma ray flux has an almost flat shape in the radial direction within the corner gap, this error is caused by the core axial component of the TIP measurement in that string ERMZ ( K
, L) almost match the correct one.

On the other hand, in estimating the best TIP measurement value, the core radial component ERCR (
L) and its core axial component ERCZ (K, L)
I don't take it. The reason is that the power distribution calculated by the built-in physical model of the power distribution calculation f1 header 8 is such that even if the 81f3 accuracy of the core axial component is insufficient, the core radial component is calculated with high accuracy. This is because it has been experimentally confirmed that In other words, in Figure 4, the horizontal axis is the distance from the center of the core to the fuel assembly, and the vertical axis is the error in the average La140 concentration in each fuel assembly. Equivalent La converted from the needle cylinder power distribution by the physical model
The error of 1401 degrees is shown by points. As can be seen from the figure, these errors pose no problem in practice and can be completely ignored. This means that the core radial component of the calculated power distribution dew value based on the physical model has good accuracy, and at the same time, the linear transformation for the power distribution (
This confirms that the core radial direction component ERCR(L) of the calculated value for the TIP measurement value excluded by the formula [[) has good accuracy.

For the reasons mentioned above, when estimating the best TIP3111 constant value, take the core (\ direction component ERCR (+-)) from the gIn value for the TIP measurement value, and calculate the core axial direction component ERCZ (K, L). Let's say Tora rZ.

Next, the best output distribution estimating device 9 corrects the calculated output distribution value using the following equations (XI) and (Xn) so that the output density of the node matches the best TIP measurement value estimated above. In other words, G (K, I, J) = ERB (K, L)/ERC (K, L)...
...(XI) PC(K, I, J) = G (K, 1.J)・P(K, I, J)...
...(XII) Here, the node (K, 1.,)) is the measurement position (K.

Refers to everything adjacent to L). Also, PC (K, I, J) which is ringed by the formula (XII) is expressed as P(
K.1. J), the correction factor G(K,
1. As is clear from the definition of J) and formula (III),
The best TIP measurement value ER[3(K
, L) is obtained. In this sense, PC(K, I, J) is the best T
It is said to comply with IP measurements.

In the 1d model, the best output distribution estimation device 9 uses the P obtained above.
C(K, T, J) is corrected so that its radial component matches that of the output distribution calculation value based on the physical model. That is, PR (1, J) ...... (Xl [I) PZ (K, I, J) = P (K, I, J) / PR (r, J) ... (XIV
), P(K, I, J) is converted into core axial component PZ(K,
1. J) and core radial component PR(1, J).

Here, considering that the accuracy of the core radial component of the power distribution calculated by the physical model is high as described above, the above PC (
The radial components of K, I, J) are corrected using the following equation (XV) so that they match PR(T, J).

Here, PB (K) calculated by the above formula (XV).

1, J) is what is finally determined by the best output distribution estimating device 9 of the present invention, and is estimated by extracting only highly accurate components from each of the TIP output signal and the calculated value of the physical model. This is called the best output distribution.

In addition, in the above example, a case is described in which measurements are performed using TIP, which is an in-core moving gamma-ray flux measuring instrument, but L
In the case of PPM, the measured value of TIP is determined as follows, and the process described above is performed exactly as is.

That is, in FIG. 3, LPP at the same measurement point
The two-sided constant values of M3 and TIP4 do not match due to differences in measurement targets, differences in detection efficiency, etc.

However, the coefficient that avoids matching the two is Lp R
M3 can be determined for each measurement point during measurement using a mobile gamma-ray flux measuring device. Therefore, the measured value of gamma ray flux at each measurement setting point at each measurement point is the value obtained by multiplying the measurement value of the thermal neutron flux measuring device at the current point in the corner gap by this coefficient, and the value of the thermal neutron flux measuring device It can be determined by interpolating the difference between the measured value of the gamma-ray flux measuring device at the closest point in time to the present time at the position where the gamma-ray flux measuring device is placed, and adding it to the measured value of the gamma-ray flux measuring device.

[Effect of Brahma] As explained above, the online power distribution monitoring of the nuclear reactor of the present invention 8! According to the device, the power distribution within the reactor can be monitored online with high accuracy, so the operating margin can be reduced.

Therefore, the efficiency of nuclear reactor operation can be further improved compared to conventional devices.

[Brief explanation of the drawing]

FIG. 1 is a block configuration diagram of an embodiment of the present invention, FIG. 2 is a τ1 view showing the arrangement of an in-core gamma-ray flux measuring device to which the present invention is applied, and FIG. 3 is an inside of the core in FIG. FIG. 4 is a diagram for explaining the deviation of the gamma-ray flux measuring device, and is a diagram for explaining the measured viscosity of the core radial component of the power distribution measured by the physical model according to the present invention. 3...Fixed neutron measuring instrument 4...Variable α-ray flux measuring instrument 5...Measuring instrument for measuring core current data 6...Output distribution monitoring device 7...Data sampler 8... Output distribution monitoring device 9... Best output TFI estimating device 10... Input/output device (8733) Agent Patent attorney Akira Inomata and others 1
Figure 2

Claims (1)

[Claims] A core current data measuring device, a power distribution calculation device that inputs the measured values from the core current data measuring device and calculates a power distribution in the reactor core based on a pre-built-in physical model, and an in-core gamma ray flux measuring device; The measured value of this in-core gamma-ray flux measuring device is input and only its core axial direction component is extracted, and the output distribution calculated by the power distribution calculation device is input and the obtained core axial direction component is extracted. 1. An online power distribution monitoring device for a nuclear reactor, comprising: a best power distribution estimating device that estimates the best power distribution by correcting the measured value of a gamma ray flux measuring device using a component in the core axis direction.