JP2023075550A - Method of measuring power distribution in nuclear reactor and device therefor - Google Patents

Abstract

To enable measurement of a power distribution of a nuclear reactor having limitation on locations each inserted with a neutron detector, like high temperature gas reactors and fast reactors.SOLUTION: Provided is a method of measuring a power distribution at a reactor center on the basis of a neutron transport equation which expresses a relationship among power densities of a plurality of fuel elements in a pressure vessel of a nuclear reactor, output signals from neutron detectors at positions of a plurality of neutron detectors inside and outside the pressure vessel, and detector sensitivities related to positions of the fuel elements and the neutron detectors. In the method, a power distribution at the reactor center of the nuclear reactor is calculated from a product of a pseudo inverse matrix related to the detector sensitivities and an output signal matrix from the neutron detectors.SELECTED DRAWING: Figure 3

Description

The present invention relates to a method and apparatus for measuring power distribution in a nuclear reactor by inverse solution using neutron detector signals.

In current light water reactors, the environment inside the reactor is about 300 to 400°C. burnup management is performed (Patent Document 1). On the other hand, since the temperature inside the HTGR reaches about 1000°C at maximum and about 600°C for the fast reactor, no attempt has been made to measure the power distribution inside the reactor.

Japanese Patent No. 5954902

In recent years, ceramic detectors and other high-temperature resistant neutron detectors have been developed. There are limited places to insert the detector inside. Also, from the viewpoint of the reactor internals, it seems realistic to insert the detector into the reactor both in high-temperature gas-cooled reactors and in fast reactors, by using control rod guide tubes together. It is impractical to place the detector directly on the

As described above, the present invention is applicable to nuclear reactors such as high-temperature gas reactors and fast reactors, which have long neutron ranges and cannot insert detectors into the reactors, or whose insertion locations are limited even if possible. The main purpose is to enable measurement of in-core power distribution.

As described above, the main object of the present invention is to make it possible to measure the in-reactor power distribution of nuclear reactors such as high-temperature gas reactors and fast reactors, in which the places where neutron detectors are inserted are limited. However, as a matter of course, it goes without saying that the method can also be applied as it is to light water reactors such as boiling water reactors and pressurized water reactors, in which the neutron detector can be inserted relatively freely.

A method for measuring a power distribution in a nuclear reactor according to a first aspect of the present invention comprises power densities of a plurality of fuel elements within a pressure vessel of a nuclear reactor and power densities from neutron detectors at positions of the plurality of neutron detectors outside the pressure vessel. A method for measuring a core power distribution based on a neutron transport equation representing the relationship between output signals and detector sensitivities with respect to the positions of the fuel elements and the neutron detectors, the method comprising: The power distribution of the core of the nuclear reactor is calculated from the product of the matrix and a pseudo inverse matrix relating to the detector sensitivity.

More specifically, the method for measuring the power distribution in the reactor of the present invention is based on the power densities pi of n multiple fuel elements i inside the pressure vessel and the neutrons at m multiple detector positions j inside and outside the pressure vessel A method for measuring the power distribution of a core according to the following equation based on the neutron transport equation representing the relationship between detector signal Rj and detector sensitivity _wj , _i for said fuel element i and said detector position j, comprising:
(1) a first step of obtaining the neutron detector signal from the following equation;

(2) a second step of representing the neutron detector signal _Rj , the power density pi, and the detector sensitivity _wj , _i in a matrix representation of the following equation;

(3) a third step of calculating the pseudo-inverse matrix W ⁺ of the matrix representation of the detector sensitivity w _j , _i by the matrix representation of the following equation;

(4) A fourth step of calculating the matrix representation of the power density pi by the matrix representation of the neutron detector signal R _j and the pseudo inverse matrix W ⁺ by the following formula:

is characterized by measuring the power distribution in the reactor by sequentially executing

A nuclear reactor power distribution measuring apparatus according to a second aspect of the present invention comprises power densities of a plurality of fuel elements in a reactor pressure vessel, neutron detector signals at a plurality of detector positions inside and outside the pressure vessel, A device for measuring the power distribution of the reactor core based on a neutron transport equation representing the relationship between the detector sensitivity with respect to the position of the fuel element and the detector, and the core of the nuclear reactor by inverse analysis of the detector sensitivity A storage device storing a program indicating a procedure for calculating the output distribution of the detector, and an arithmetic device for inputting a signal from the detector and performing a predetermined calculation based on the program.

More specifically, the reactor power distribution measurement apparatus of the present invention measures the power densities _pi of n multiple fuel elements i inside the pressure vessel and the m multiple detector positions j inside and outside the pressure vessel. A device for measuring the power distribution of the core by the following equation based on the neutron transport equation representing the relationship between the neutron detector signal _Rj and the detector sensitivity _wj , _i with respect to the fuel element i and the detector position j. and the program is
(1) a first step of obtaining the neutron detector signal from the following equation;

(2) a second step of representing the neutron detector _{signal Rj, the power density pi, and the detector sensitivity wj,i in a matrix} representation of the following equation;

(3) a third step of calculating the pseudo-inverse matrix W ⁺ of the matrix representation of the detector sensitivity w _j , _i by the matrix representation of the following equation;

(4) A fourth step of calculating the matrix representation of the power density _pi by the matrix representation of the neutron detector signal _Rj and the pseudo-inverse matrix W ⁺ by the following formula:

are sequentially executed.

The present invention achieves the following effects.
(1) For a small core with a long range of neutrons, it is possible to measure the in-core power distribution using detector signals from only the out-of-core detectors.
(2) If the detector can be inserted into the reactor and the insertion location is limited, it is possible to measure the in-core power distribution of the entire core from the detector signals at the limited location in the reactor.

As described above, even in a light water reactor with a short range of neutrons, the power distribution in the reactor can be measured by applying the present invention, and the resolution of the power distribution measurement can be improved more than before. is possible.

Explanatory drawing of the detection sensitivity of an ex-core detector of a light water reactor. (A) is an explanatory diagram of the detection sensitivity of an ex-core detector in a high-temperature gas-cooled reactor, and (B) is an explanatory diagram of the detection sensitivity of an in-core detector. 4 is a flow chart schematically showing a method of measuring the power distribution in the reactor according to the present invention; FIG. 2 is a schematic explanatory diagram showing the positional relationship between an out-of-core detector used to measure the in-reactor power distribution according to the present invention and the core. The figure which shows the driving example of the out-of-core detector in a high temperature gas-cooled reactor. The figure which shows the driving example of the in-core detector in a high temperature gas-cooled reactor.

As shown in Fig. 1, the detection sensitivity from each fuel assembly to the out-of-core detectors of a light water reactor is limited to the fuel bodies in the outer periphery of the core, and the output of the in-core position in the core of the reactor by the out-of-core instrumentation. A direct measurement of the distribution is impossible in principle. In a pressurized water reactor (PWR), for example, as disclosed in Japanese Patent No. 5954902, monitoring of the power deviation (axial offset) in the axial direction of the power distribution to control the Xe oscillation Therefore, the axial offset is evaluated by ex-core instrumentation. Although the purpose is to evaluate the power distribution, it is actually a problem of determining the absolute value of the assumed power distribution from the measured values at the periphery, and the power distribution itself at the center of the core is assumed, which is not the purpose of the present invention. Such a detailed power distribution as used for burnup control cannot be obtained.

This is because light water has excellent performance as a moderator in light water reactors, so neutrons generated by nuclear fission collide with hydrogen contained in light water and instantly change into thermal neutrons, which are easily absorbed by nuclear fuel materials. This is because the range is very short and, as shown in FIG. 1, neutrons generated in the center of the core hardly leak out of the reactor.

On the other hand, Fig. 2 shows the detector sensitivities when assuming the arrangement of the detectors outside and inside the reactor as assumed in the present invention, targeting the system of the test and research reactor HTTR of the high-temperature gas-cooled reactor. showing. It can be seen that the detector sensitivity is distributed over a wider range than that of the light water reactor. Assuming this characteristic, we present a technique for inverse analysis of the detector signal and reconstruction of the output distribution. The measurement principle is as follows. As shown in FIG. 2, the neutron sensitivity from the detector position to the nuclear fuel region is evaluated in advance by neutron transport calculations.

w(r,r _d ) is the sensitivity of the detector to detect neutrons generated at position r in the reactor from neutrons generated at detector position _rd , and S(r) is the fission neutron source. On the other hand, in order to evaluate the detector sensitivity, it is necessary to solve the following neutron transport equation.

L is a neutron deficiency operator, which represents neutron deficiency due to neutron transport, scattering, and absorption. As a result, the detector response of the obtained neutron flux φ(r) becomes the detector signal.

_Σd indicates the cross-sectional area of the neutron reactant in the detector. R(r _d , r) is the detector signal at detector position _rd due to neutrons generated at in-core position r.

Here, if the neutron source S(r) is assumed to be a unit fission source from only the in-core position r _i as follows, (where χ is the fission spectrum)

Since this detector rate is itself the detector sensitivity,

becomes. On the other hand, for convenience, since the power distribution is evaluated for a fuel element with a specific breadth, the detector sensitivity should also be defined for the fuel element, defined as:

This is the sensitivity of the neutrons emitted from the fuel element i when the detector at the detector position j detects neutrons.

On the other hand, when obtaining this detector sensitivity, the neutron transport calculation by Eq. However, there is a risk of an increase in calculation time and the inclusion of errors due to a large number of numerical processings. Generally, a solution method that solves the following adjoint neutron transport equation is used.

L ⁺ is the adjoint neutron deficiency operator. Accordingly, the detector sensitivity of equation (5) can be expressed as follows.

In this calculation method, the sensitivity of the detector to each fuel element can be evaluated by solving the adjoint neutron transport equation (7) only once and by performing the tabulation work of the equations (6) and (8).

where the detector signal R _j at detector position j is expressed by the sensitivity w _j from fuel element _i and the power density _pi for fuel element i at the detector at detector position j: can be done. To clarify the physical image, the fission neutron source is replaced by the proportional power density.

Obtaining detector signals at multiple detector positions j yields multiple equations, which in matrix representation can be expressed as:

If n is the number of fuel elements and m is the number of detector positions, the detector sensitivity is an m-by-n matrix. If the number of fuel elements and the number of detector positions are the same, n=m, and the row vectors of the detector sensitivities are linearly independent, the detector sensitivity matrix W becomes a regular matrix of n rows and n columns, and the inverse matrix is known to have

In this case, the vector P of power densities can be obtained directly from equation (11) as an exact solution. This is synonymous with the theorem that the number of equations equal to the number of variables is required as a condition for obtaining the solution of simultaneous linear equations. However, in reality, the linear independence of the detection efficiency of n measurement points measured arbitrarily is not guaranteed, and even if n measurement points are set, m as a linearly independent measurement point is m It is quite possible that <n.

On the other hand, finding the least-squares solution gives a general expression that includes this exact solution.

The condition for minimizing the squared error J defined by equation (12) is

is the solution of This solution is

is obtained as

The steps up to this point are briefly illustrated in FIG. where W ⁺ is called the pseudo-inverse. As described above, W is an m-by-n matrix, which does not have an inverse unless n=m and the row vectors of the detector sensitivities are linearly independent. If this condition is met,

By the same calculation, an exact solution similar to the formula (11) can be obtained.
On the other hand, when n≠m, the solution obtained by Equation (14) is said to be a least squares solution but not unique, that is, it is not guaranteed to be a true value.

Here, if m<n, that is, if the number of equations is insufficient for the number of variables (here, the number of measurement points is insufficient), the solution as a simultaneous linear equation is determined. Although it is not possible, it is possible to determine the solution as an approximate solution that minimizes the squared error shown in equation (13). However, as will be confirmed in later numerical experiments, an error occurs in the predicted output distribution due to the lack of measurement points.

On the contrary, when m>n, i.e., when there are more equations than the number of variables (here, when the number of measurement points is more), it is still a least-squares solution and is considered non-unique. ing. However, as will be confirmed in later numerical experiments, in this problem, exact solutions are obtained even when an excessive number of measurement points are set. This is self-evident from the following point of view. This is because if the number of equations is large, the exact solution can be obtained using the equation (16) by ignoring the number of equations.

For example, if the detector sensitivities at m measurement points are linearly independent, the measurement points (equations) are divided into n groups and mn groups. First, the exact solution of the output distribution is obtained from the equations (14) and (16) for n groups. Of course, the detector signals of the remaining mn groups are the detector signals obtained from equation (9), that is, the detector signals obtained from the strict output distribution (actual output distribution), even if Using the exact output distributions obtained from equations (14) and (16) in n groups, in equation (9), even if the remaining mn groups of detector signals are used, both matches. In other words, in order to correctly calculate the output distribution, it is sufficient if m≧n.

In addition, if some error is mixed in the measurement,

If the error has a uniform distribution such as electrical noise, considering the properties of the least-squares method, the more measurement points, the more likely it is to eliminate the influence of uncertainty such as noise. This tendency can also be confirmed from the fact that the more measurement points there are, the more the errors that are thought to be numerical errors are reduced in the numerical experiments that will be carried out later. When considering noise explicitly, it can be said that the larger the number of measurement points, the better.

In the present invention, many measurement points can be secured by moving the detector, as described in detail below. However, due to constraints such as the structure of the installed nuclear reactor and material integrity such as high temperature and high dose environment, there are cases where sufficient measurement points cannot be secured, or the resolution of the power distribution to be evaluated is in units of fuel rods. For this reason, it is quite conceivable that the number of measurement points is insufficient for the number of fuel elements that have a power distribution. In that case, the validity of the measurement results shall be guaranteed by the following method. Conveniently, the power distribution can be determined by using the equation (14) by calculating the pseudo inverse matrix shown in the equation (15) regardless of the magnitude relationship between the number of measurement points m and the number of fuel elements n. Whether or not the number of measurement points is sufficient can be determined by checking the rank of this pseudo-inverse matrix. The rank of a matrix indicates the number of linearly independent rows in each row of the matrix. The pseudo-inverse has n rows and n columns, and the maximum rank is n.

As the number of measurement points increases, the rank increases and becomes saturated at the maximum value of n. Numerical experiments associated with the present invention have confirmed that an exact solution can be obtained when the order takes the maximum value of n. On the other hand, if the rank is less than n, an error occurs in the predicted power density. According to the present invention, in the numerical experiments described later, from the results shown in Table 2, it is possible to reproduce the power density distribution with sufficiently acceptable accuracy when the rank of the pseudo-inverse matrix is 0.8n or more. is confirmed.

Therefore, the number of measurement points should be equal to or greater than the number of fuel elements. The number of fuel elements depends not only on the number of fuel assemblies but also on the required resolution. If it is desired to further divide the inside of the fuel assembly for burnup control, the number of fuel elements is further increased. Taking the HTTR, which is a test and research reactor of a high-temperature gas-cooled reactor, as an example, it is composed of 150 hexagonal prism-shaped fuel blocks with a distance between faces of 36 cm and a height of 58 cm. Even if 150 detector signals are measured at the minimum, it is not realistic to measure with individual detectors, and it is necessary to move a single detector or multiple detectors to measure detector signals. .

Therefore, in this method, it is assumed that detector signals are measured by assuming external instrumentation, in-core instrumentation, and, if necessary, a combination of both.

When measuring the power distribution using only the detector outside the reactor, as shown in FIG. Driving the detector along a spiral trajectory 10 to obtain a signal is the most ideal way to use this technique. With this method, it is possible to perform measurements with a series of movements using only one detector, and this is a drive method that is also used in general X-ray CT. However, from the standpoint of manufacturability and maintainability, it is simpler to arrange a plurality of detectors on the outer circumference of the pressure vessel and vertically drive each detector. This makes it possible to measure from an almost infinite number of measurement points.

When adopting for in-reactor instrumentation, in the case of high-temperature gas-cooled reactors, the environment inside the reactor is extremely high, and in the case of fast reactors, radiation damage to detectors is conspicuous. Instead of placing them in the core, it is practical to use control rod guide tubes, insert them when in use, and store them outside the reactor when not in use. Since the control rod guide tubes exist independently of the fuel in both the HTGR and the fast reactor, there is room for the control rod guide tubes, and it is possible to change the design to drive the detector.

Figure 6 shows an example of a high temperature gas reactor. Assume the insertion of the detector from the standpipe 20 that houses the control rods. When not in use, the detector is retracted and stored within the standpipe 20 . It can be stored at a temperature of about 300°C by a cooling flow, and radiation damage can be prevented. When in use, insert continuously to secure the number of measurement points. In the case of the HTTR, 30 fuel bodies are arranged in the radial direction, whereas there are 16 control rod guide blocks. In the case of evaluation, if two measurement points are provided in the height direction within a fuel block height of 58 cm at intervals of 29 cm, the same number of measurement points as the fuel block can be secured. When avoiding high-temperature regions as shown in FIG. 4, the number of measurement points is reduced, so it is expected that the accuracy of output distribution evaluation will be reduced. If the measurement accuracy is unacceptable, it is possible to avoid a decrease in accuracy by taking countermeasures such as increasing the number of in-core detectors by changing the design.

In addition, the higher the number of measurement points, the higher the resolution of the output distribution to be measured. Become.

In the process of conceiving the present invention, a core with a long neutron range, which is different from light water reactors, was assumed. I can expect it. In the in-core instrumentation of light water reactors, the power distribution that can be predicted from the measured neutrons is treated as representing the average power of the fuel assembly around the detector. Considering that we are measuring leaked neutrons, we can directly determine the power distribution for each ensemble by performing inverse analysis using the pre-evaluated detector sensitivity distribution.

Electric power companies operating current light water reactors monitor the burnup directly and manage the fuel in the reactor by directly measuring the power distribution using in-core instrumentation. On the other hand, for HTGRs and fast reactors, no in-core instrumentation has been developed due to the high temperature environment, and direct in-core power distribution measurement technology has not been developed. The present invention makes it possible for the first time to measure the in-core power distribution of both advanced reactors, enabling the same in-core fuel management as in light-water reactors, and significantly improving safety.

In a small core system, if sufficient detector sensitivity can be secured up to the fuel bodies in the center of the core, it is possible to measure the power distribution using only external instrumentation. By driving the out-of-core detectors, the number of detectors required for obtaining the necessary measurement points can be reduced, and the trouble of correcting the detection efficiency due to the construction of a plurality of detectors can be avoided.

In the core where the in-core instrumentation can be installed, it is possible to measure the power distribution in a wide range for the limited insertion position in the core. It is possible to measure the output distribution even if the place is limited. As for the in-core instrumentation, by making the in-core detector movable, it is possible not only to increase the number of measurement points but also to avoid deterioration when the detector is not used. In high-temperature gas-cooled reactors and fast reactors, the control rod guide tubes are independent from the fuel assemblies, and there is plenty of room for locating driven detectors.

The performance of the present invention was applied to the high temperature gas reactor test reactor HTTR system and confirmed by simulation. As for the external instrumentation, Table 1 shows the results of measuring the power distribution of 150 fuel blocks at a total of 180 points, 36 points on the circumference of the pressure vessel and 5 points on the height.

The average measurement error of all fuel blocks was 1.2×10 ⁻⁹ %, and the block showing the maximum error showed a negligible error of about 1.1×10 ⁻⁸ %. The in-core instrumentation was evaluated by measuring 320 points including 16 detectors and 20 height positions. Table 1 also shows the results. The average measurement error of all the fuel blocks is 3.2×10 ⁻¹¹ %, and the block showing the maximum error is about 2.4×10 ⁻¹⁰ %. Table 2 shows the results when partial insertion was performed in order to avoid insertion of the detector into the lower part of the core where the temperature becomes high.

The positions of the measurement points are fixed, and the measurement points below the core are excluded from the 20 points in the height direction by partial insertion. Even in the case of 50% insertion, it shows a good match that can be said to be an exact solution. At 45% and 40% partial insertion, core average errors of 0.28% and 1.1%, and maximum local errors of 5.0% and 6.5% are significant. As is clear from the reduction in the rank of the inverse matrix, this error is due to the lack of measurement points rather than the position of partial insertion. There is room for improvement to eliminate. In this way, the power distribution of the entire core can be measured by inserting about half of the detectors into the reactor, and it is possible to avoid inserting detectors into high-temperature regions. With 50% partial insertion, the environmental temperature of the detector can be expected to drop by about 200° C., and from the viewpoint of heat resistance of the detector, the load applied to the detector can be greatly reduced.

The applicability of the power distribution measurement method using ex-core instrumentation is determined by the relationship between the neutron range and the core size. Table 3 shows the average travel distance from generation to annihilation of neutrons as the range of neutrons for each reactor type.

Diffusion distances were evaluated and compared to the fuel section. While the light water reactor is about 7 cm, the high temperature gas reactor is about 4 times that, and the fast reactor is about 6 times that. It is possible to estimate the power distribution of the fuel assembly over a wide range from the measured neutron information. is. The detector sensitivity of light water reactors is about 40-60 cm, as shown in FIG. Even the design of NuScale Power, LLC of the United States, which is being developed as a small nuclear reactor (SMR), cannot be applied to a light water reactor because the core radius is about 120 cm. For the HTGR, a wide detector sensitivity distribution is obtained as shown in FIG. Assuming that the HTTR effective core radius of about 130 cm can be observed from the outer circumference of the core, the higher power of the high-temperature gas-cooled reactor will increase the power density and increase the length of the core. From the point of view of safety, since it takes the shape of an annular core without arranging fuel in the center of the core, not only HTTR with a core output of 30 MW, but also a 50 MW commercial reactor design designed by the Japan Atomic Energy Agency , 165 MW core, and 600 MW core, and is expected to be applied to most of the designs of high-temperature gas-cooled reactors classified as SMR (Small? Modular Reactor). For fast reactors, a wider range of detector sensitivities can be expected due to the longer diffusion lengths. Even if it is possible to observe a fuel width of about 130 cm, which is equivalent to that of a high-temperature gas-cooled reactor, the PRISM (Power Reactor Innovative Small Module) reactor, which is a typical design of fast reactor SMR, has a core radius of about 130 cm. Fully applicable. On the other hand, in a large fast reactor with a core radius of about 300 cm, it is considered difficult to apply only an external detector.

The method using in-core instrumentation can be applied not only to high-temperature gas-cooled reactors and fast reactors, but also to data processing of current in-core instrumentation in current light water reactors. By applying the core inverse analysis method, it is possible to improve the resolution of the in-core power distribution. While the current method is an integral approach in which the measured value of the in-core instrumentation is the average value of the peripheral fuel assembly power, the present analysis is a differential approach in which measurement signals are assigned to each region and reconstructed. Since it is an approach, it is a self-evident result.

In the above explanation, a pseudo-inverse matrix is used to show the essential mathematical structure, and the rank of the pseudo-inverse matrix is used as an index for measuring the linear independent component. However, there are also approaches such as numerical solutions that do not use inverse matrices and maximum likelihood methods that replace the least squares method, and there are multiple implementations of inverse methods.

10... Spiral orbit 20... Standpipe

Claims (9)

power densities of multiple fuel elements within a reactor pressure vessel; output signals from neutron detectors at multiple neutron detector locations outside the pressure vessel; and detector sensitivity with respect to locations of said fuel elements and said neutron detectors. A method for measuring the power distribution of a core based on the neutron transport equation representing the relationship of
A method for measuring a power distribution in a nuclear reactor, wherein a power distribution in a core of the nuclear reactor is calculated from a product of a matrix of output signals from the neutron detectors and a pseudo-inverse matrix relating to the sensitivity of the detectors. 2. The method for measuring the power distribution in a nuclear reactor according to claim 1, wherein _pi is the power density of n multiple fuel elements i inside the pressure vessel, and m multiple detector positions j inside and outside the pressure vessel detect neutrons. Let R _j be the detector signal, w _j , _i be the detector sensitivity for said fuel element i and said detector position j, and W ⁺ be the pseudo-inverse matrix for detector sensitivity,
The output signal from the neutron detector is

and the pseudo-inverse matrix for the detector sensitivity is

A method for measuring power distribution in a nuclear reactor, characterized in that the calculation is performed by matrix representation of In the method for measuring the power distribution in the reactor according to claim 1,
power densities pi of n multiple fuel elements i inside the pressure vessel, neutron detector signals Rj at m multiple detector positions j inside and outside the pressure vessel, detectors for said fuel elements i and said detector positions j A method for measuring the power distribution of the core by the following equation based on the neutron transport equation representing the relationship between sensitivities w _j and _i ,
(1) a first step of obtaining the neutron detector signal from the following equation;

(2) a second step of representing the neutron detector _{signal Rj, the power density pi, and the detector sensitivity wj,i in a matrix} representation of the following equation;

(3) a third step of calculating the pseudo-inverse matrix W ⁺ of the matrix representation of the detector sensitivity w _j , _i by the matrix representation of the following equation;

(4) A fourth step of calculating the matrix representation of the power density _pi by the matrix representation of the neutron detector signal _Rj and the pseudo-inverse matrix W ⁺ by the following formula:

A method for measuring the power distribution in a nuclear reactor, characterized by sequentially executing In the method for measuring the power distribution in the reactor according to claim 1 or 2,
A method for measuring power distribution in a nuclear reactor, wherein the neutron detector signals are acquired by driving neutron detectors installed at a plurality of detector positions inside and outside the pressure vessel. In the method for measuring the power distribution in the reactor according to any one of claims 1 to 3,
A method for measuring power distribution in a nuclear reactor, wherein a measured value is output as a valid measurement result when the rank of the pseudo-inverse matrix is 0.8n (n is the number of the plurality of fuel elements) or more. 1 represents the relationship between the power density of multiple fuel elements within a reactor pressure vessel, neutron detector signals at multiple detector positions inside and outside the pressure vessel, and detector sensitivity with respect to the fuel element and detector positions. A device for measuring the power distribution of a core based on the neutron transport equation,
A storage device storing a program showing a procedure for calculating the power distribution of the core of the nuclear reactor by inverse analysis of the detector sensitivity, and a predetermined calculation based on the program by inputting the signal from the detector. A power distribution measuring device in a nuclear reactor, characterized by comprising an arithmetic device for performing In the reactor power distribution measurement device according to claim 6,
for the power density p _i of n multiple fuel elements i inside the pressure vessel, the neutron detector signals R _j at m multiple detector positions j inside and outside the pressure vessel, and for said fuel element i and said detector position j A device for measuring the power distribution of the core by the following equation based on the neutron transport equation representing the relationship with the detector sensitivity w _j , _i, wherein the program is:
(1) a first step of obtaining the neutron detector signal from the following equation;

(2) a second step of representing the neutron detector _{signal Rj, the power density pi, and the detector sensitivity wj,i in a matrix} representation of the following equation;

(3) a third step of calculating a pseudo-inverse matrix W+ of the matrix representation of the detector sensitivities w _j , _i by the matrix representation of the following equation;

(4) A fourth step of calculating the matrix representation of the power density _pi by the matrix representation of the neutron detector signal _Rj and the pseudo-inverse matrix W ⁺ by the following formula:

A power distribution measurement device in a nuclear reactor, characterized in that it sequentially executes. In the reactor power distribution measuring device according to claim 6 or 7,
An apparatus for measuring power distribution in a nuclear reactor, comprising means for acquiring neutron detector signals by driving neutron detectors installed at multiple detector positions inside and outside the pressure vessel. In the reactor power distribution measuring device according to any one of claims 6 to 8,
Power distribution in the reactor, characterized by comprising means for outputting a measured value as a valid measurement result when the rank of the pseudo-inverse matrix is 0.8n (n is the number of the plurality of fuel elements) or more. measuring device.

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