Method and apparatus for measuring the power distribution inside a nuclear reactor

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

[Technical Field]

[0001] This invention relates to a method and apparatus for measuring the power distribution inside a nuclear reactor by an inverse method using neutron detector signals. [Background technology]

[0002] In current light water reactors, the internal environment is around 300-400°C, so neutron detectors are directly mounted near the fuel assemblies in the reactor to measure the power distribution, and the burnup of the fuel inside the reactor is managed based on the measured power distribution (Patent Document 1). On the other hand, in high-temperature gas reactors, the internal temperature reaches a maximum of around 1000°C, and in fast reactors, it reaches around 600°C, so no attempts have been made to measure the power distribution inside the reactor. [Prior art documents] [Patent Documents]

[0003] [Patent Document 1] Patent No. 5954902 [Overview of the project] [Problems that the invention aims to solve]

[0004] In recent years, high-temperature resistant neutron detectors, such as ceramic detectors, have been developed. However, the placement of in-core instrumentation in high-temperature gas reactors is not practical, as insertion near the reactor coolant outlet is not feasible, limiting the locations where detectors can be placed inside the reactor. Furthermore, from the perspective of in-core structures, it is considered more practical to use control rod guide tubes in conjunction with detector placement in both high-temperature gas reactors and fast reactors, making it impractical to directly place detectors near each fuel assembly.

[0005] As described above, the main objective of the present invention is to enable the measurement of the in-core power distribution in reactors with long neutron ranges, such as high-temperature gas reactors and fast reactors, where it is not possible to insert detectors into the reactor, or where the insertion locations are limited.

[0006] As described above, the primary objective of the present invention is to enable the measurement of the in-reactor power distribution in reactors where the insertion location of neutron detectors is limited, such as high-temperature gas reactors and fast reactors. However, as a natural consequence, it goes without saying that the invention can also be applied to light water reactors such as boiling water reactors and pressurized water reactors, where the insertion location of neutron detectors is relatively flexible. [Means for solving the problem]

[0007] A method for measuring the power distribution inside a nuclear reactor according to a first aspect of the present invention is a method for measuring the power distribution inside a reactor core based on a neutron transport equation that represents the relationship between the power density of multiple fuel elements inside the pressure vessel of a nuclear reactor, the power signals from neutron detectors at the positions of multiple neutron detectors outside the pressure vessel, and the detector sensitivity relating to the positions of the fuel elements and the neutron detectors, characterized in that the power distribution inside the reactor core is calculated from the product of a matrix of power signals from the neutron detectors and a pseudo-inverse matrix relating to the detector sensitivity.

[0008] More specifically, the method for measuring the power distribution inside a nuclear reactor according to the present invention comprises the power density pi of n multiple fuel elements i inside a pressure vessel, the neutron detector signal Rj of m multiple detector positions j inside and outside the pressure vessel, and the detector sensitivity w related to the fuel elements i and the detector positions j. j , i A method for measuring the power distribution of the reactor core by the following equation based on the neutron transport equation that expresses the relationship between, (1) The first step is to determine the neutron detector signal from the following formula, TIFF0007894618000001.tif14170

[0009] (2) The neutron detector signal R j The output density pi and the detector sensitivity w j , i The second step is to represent and in matrix form as shown in the following equation, TIFF0007894618000002.tif14170

[0010] (3) The detector sensitivity w j , i The pseudo-inverse matrix W of the matrix representation of + is calculated by the matrix representation of the following formula in the third step, TIFF0007894618000003.tif14170

[0011] (4) The matrix representation of the output density pi is calculated by the following formula using the matrix representation of the neutron detector signal R j and the pseudo-inverse matrix W + in the fourth step, TIFF0007894618000004.tif12170

[0012] By sequentially executing the above steps, it is characterized by measuring the output distribution in the nuclear reactor.

[0013] The device for measuring the output distribution in a nuclear reactor according to the second aspect of the present invention is a device for measuring the output distribution of the core based on the neutron transport equation representing the relationship between the output density of a plurality of fuel elements in the pressure vessel of the nuclear reactor, the neutron detector signal at the positions of a plurality of detectors inside and outside the pressure vessel, and the detector sensitivity regarding the positions of the fuel elements and the detectors, and includes a storage device storing a program showing a procedure for calculating the output distribution of the core of the nuclear reactor by inverse analysis of the detector sensitivity, and an arithmetic device for inputting a signal from the detector and performing a predetermined calculation based on the program.

[0014] More specifically, the device for measuring the output distribution in a nuclear reactor according to the present invention is a device for measuring the output distribution of the core by the following formula based on the neutron transport equation representing the relationship between the output density p i of n plurality of fuel elements i in the pressure vessel, the neutron detector signal R j at m plurality of detector positions j inside and outside the pressure vessel, and the detector sensitivity w j , i regarding the fuel element i and the detector position j. The program includes (1) The first step of obtaining the neutron detector signal from the following formula, TIFF0007894618000005.tif14170

[0015] (2) The neutron detector signal R j And the output density p i And the detector sensitivity w j , i The second step is to represent and in matrix form as shown in the following equation, TIFF0007894618000006.tif14170

[0016] (3) The detector sensitivity w j , i The pseudo-inverse matrix W of the matrix representation + The third step is to calculate it using the matrix representation of the following formula, TIFF0007894618000007.tif14170

[0017] (4) The output density p i The matrix representation of the neutron detector signal R j The matrix representation of and the pseudo-inverse matrix W + The fourth step is calculated using the following formula and TIFF0007894618000008.tif12170

[0018] This method is characterized by sequentially executing the following steps. [Effects of the Invention]

[0019] The present invention achieves the following effects. (1) For small reactor cores with long neutron ranges, it becomes possible to measure the power distribution inside the reactor using only detector signals from external detectors. (2) If a detector can be inserted into the reactor and the insertion location is limited, it becomes possible to measure the power distribution throughout the entire reactor core from the detector signal at that limited location within the reactor.

[0020] As explained earlier, even in light water reactors where neutrons have a short range, the present invention can be applied to measure the power distribution within the reactor, and it is possible to increase the resolution of power distribution measurements compared to conventional methods. [Brief explanation of the drawing]

[0021] [Figure 1] A diagram illustrating the detection sensitivity of external detectors in light water reactors. [Figure 2] (A) is an explanatory diagram of the detection sensitivity of an external detector in a high-temperature gas reactor, and (B) is an explanatory diagram of the detection sensitivity of an internal detector. [Figure 3] A flowchart illustrating the method for measuring the power distribution inside a nuclear reactor according to the present invention. [Figure 4] A schematic diagram illustrating the positional relationship between an external detector used for measuring the power distribution inside a nuclear reactor and the reactor core according to the present invention. [Figure 5] A diagram showing an example of driving an external detector in a high-temperature gas reactor. [Figure 6] A diagram showing an example of how to operate an in-furnace detector in a high-temperature gas reactor. [Modes for carrying out the invention]

[0022] As shown in Figure 1, the detection sensitivity from each fuel assembly to external detectors in a light water reactor is limited to the fuel assemblies on the outer periphery of the core, and it is fundamentally impossible to directly measure the power distribution at the core position within the reactor using external instrumentation. In pressurized water reactors (PWRs), as disclosed in, for example, Japanese Patent No. 5954902, external instrumentation is used to evaluate the axial offset of the power distribution in the axial direction to the upper and lower parts of the core in order to control Xe oscillations. Although the purpose is to evaluate the power distribution, in reality it is a matter of determining the absolute value of the assumed power distribution using measurements from the periphery, and the power distribution in the core itself is hypothetical, so a detailed power distribution like that used for burnup management, which is the objective of this invention, cannot be obtained.

[0023] This is because light water has excellent performance as a moderator in light water reactors. Neutrons generated by nuclear fission collide with hydrogen contained in the light water and are instantly converted into thermal neutrons that are easily absorbed by the nuclear fuel material. As a result, the range of the neutrons is very short, and as shown in Figure 1, there is almost no leakage of neutrons generated in the center of the reactor core to the outside of the reactor.

[0024] On the other hand, Figure 2 shows the detector sensitivity for the High Temperature Gas Reactor (HTTR) system, assuming the arrangement of external and internal detectors as envisioned in this invention. Compared to that of a light water reactor, it can be seen that the detector sensitivity is distributed over a wide range. Assuming this characteristic, we present a technique for performing inverse analysis of the detector signal and reconstructing it into an output distribution. The measurement principle is as follows: As shown in Figure 2, the sensitivity of neutrons from the detector position to the nuclear fuel region is evaluated in advance by neutron transport calculations. TIFF0007894618000009.tif18170

[0025] w(r,r d ) Neutrons generated at position r inside the reactor, at detector position r d The sensitivity of the detector to detect neutrons generated in the region is given by S(r), where S(r) is the source of the fission neutrons. On the other hand, in order to evaluate this detector sensitivity, it is necessary to solve the following neutron transport equation. TIFF0007894618000010.tif13170

[0026] L is a neutron deficiency operator, representing neutron deficiencies due to neutron transport, scattering, and absorption. The resulting neutron flux φ(r) and the detector's reaction become the detector signal. TIFF0007894618000011.tif14170

[0027] Σ d R(r) indicates the cross-sectional area of ​​the neutron reactant in the detector. d ,r) is the position of the detector r due to neutrons generated at position r inside the reactor. d This becomes the detector signal.

[0028] Here, the neutron source S(r) is located at the reactor position r as follows. i If we assume that the unit fission source is solely from (where χ is the fission spectrum), then TIFF0007894618000012.tif18170

[0029] This detector rate itself is the detector sensitivity, TIFF0007894618000013.tif15170

[0030] This is the result. On the other hand, for convenience, the power distribution is evaluated for a fuel element with a specific width, so the detector sensitivity should also be defined for the fuel element, and is defined as follows. TIFF0007894618000014.tif20170

[0031] This is the sensitivity of the detector at detector position j when detecting neutrons generated from fuel element i.

[0032] On the other hand, when determining the detector sensitivity, it is necessary to perform neutron transport calculations using equation (2), which is expressed as equation (4) and uses a neutron source distributed only in a specific region of fuel element i, for each fuel element. This increases the calculation time and carries the risk of errors being introduced due to the large number of numerical processing steps. Therefore, generally, a solution method that solves the following associated neutron transport equation is used. TIFF0007894618000015.tif14170

[0033] L + is the adjoint neutron defect operator. Thus, the detector sensitivity in equation (5) can be expressed as follows: TIFF0007894618000016.tif17170

[0034] This calculation method allows for the evaluation of detector sensitivity for each fuel element by solving the associated neutron transport equation (7) only once and performing the aggregation of equations (6) and (8).

[0035] Here, the detector signal R at the detector position j j is the sensitivity w from the fuel element i to the detector at the detector position j j , i and the output density p for the fuel element i i can be expressed as follows. For the sake of clarity of the physical image, the fission neutron source is replaced by an output density proportional to it. TIFF0007894618000017.tif20170

[0036] By obtaining detector signals at multiple detector positions j, a plurality of equations can be obtained and can be expressed by the following equation in matrix form. TIFF0007894618000018.tif16170

[0037] When the number of fuel elements is n and the number of detector positions is m, the detector sensitivity is an m-by-n matrix. If, at the same time, the number of fuel elements and the number of detector positions are the same, n = m, and the row vectors of each detector sensitivity are linearly independent, the detector sensitivity matrix W is an n-by-n regular matrix and is known to have an inverse matrix. TIFF0007894618000019.tif15170

[0038] In this case, the vector P of the output density can be directly obtained as an exact solution from equation (11). This is synonymous with the theorem that the number of equations is required to be the same as the number of variables as a condition for obtaining the solution of a system of linear equations. However, in reality, the linear independence of the detection efficiencies at n arbitrarily measured measurement points is not guaranteed, and even if n measurement points are set, it is quite possible that m as linearly independent measurement points is m < n.

[0039] On the other hand, by obtaining the least-squares solution, a general expression including this exact solution can be obtained. TIFF00078946XXX000020.tif20170

[0040] The condition for minimizing the squared error J defined by equation (12) is Note: I assume there was a typo in ID=33 where "XXX" was in the original text. I left it as is in the translation but it should probably be a correct number. Also, I'm not sure what the significance of the ".tifxxxxxx" parts are in terms of translation, so I left them as they were. If there's more context on how to handle those, please let me know.TIFF0007894618000021.tif13170

[0041] is the solution. This solution is TIFF0007894618000022.tif14170

[0042] TIFF0007894618000023.tif16170

[0043] obtained as.

[0044] The steps up to here are briefly shown in Fig. 3. Here, W + is called a pseudo-inverse matrix. As described above, W is an m-by-n matrix, and except when n = m and the row vectors of each detector sensitivity are linearly independent, it does not have an inverse matrix. When this condition is satisfied, TIFF0007894618000024.tif13170

[0045] is obtained, and with the same calculation, an exact solution similar to equation (11) can be obtained. On the other hand, when n ≠ m, the solution obtained from equation (14) is said to be a least-squares solution but not unique, that is, there is no guarantee that it is the true value.

[0046] Here, when m < n, that is, when the number of equations is insufficient for the number of variables (here, the number of measurement points is insufficient), originally, the solution as a system of linear equations cannot be determined, but as an approximate solution that minimizes the squared error shown in equation (13), the solution can be determined. However, as will also be confirmed in subsequent numerical experiments, an error occurs in the output distribution predicted due to the shortage of measurement points.

[0047] Conversely, when m > n, that is, when there are more equations than variables (in this case, when there are many measurement points), the solution is still a least-squares solution and is not considered unique. However, as can be confirmed in later numerical experiments, an exact solution can be obtained in this problem even when an excessive number of measurement points are set. This is self-evident if we consider the following: if there are many equations, we can ignore them and obtain an exact solution using equation (16).

[0048] For example, if the detector sensitivities at m measurement points are linearly independent, we divide these measurement points (equations) into n groups and mn groups. First, we obtain the exact solution of the output distribution from equations (14) and (16) for the n groups. Naturally, the detector signals for the remaining mn groups will match even if we use the exact output distribution obtained from equation (14) and (16) for the n groups in equation (9), and then use the detector signals for the remaining mn groups in the calculation. In other words, to correctly calculate the output distribution, it is sufficient that m ≥ n.

[0049] Furthermore, if any errors are introduced during measurement, TIFF0007894618000025.tif18170

[0050] If the error has a uniform distribution, such as electrical noise, then, considering the properties of the least squares method, the more measurement points there are, the better we can expect to eliminate the effects of uncertainty such as noise. This trend can be confirmed in subsequent numerical experiments, where errors that appear to be numerical errors decrease as the number of measurement points increases. When noise is explicitly considered, it can be said that the more measurement points there are, the better.

[0051] In this invention, as will be described in detail below, many measurement points can be secured by moving the detector. However, due to constraints such as the structure of the reactor in which it is installed, the integrity of materials in high-temperature and high-dose environments, etc., it may not be possible to secure enough measurement points, or the number of measurement points may be insufficient for the number of fuel elements with power distributions, such as when the resolution of the power distribution to be evaluated is at the level of a single fuel rod. In such cases, the validity of the measurement results shall be guaranteed by the following method. Conveniently, regardless of the relative magnitudes of the number of measurement points m and the number of fuel elements n, the power distribution can be determined by using equation (14) by calculating the pseudo-inverse matrix shown in equation (15). 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 matrix is ​​n rows and n columns, and the rank is at most n.

[0052] As the number of measurement points increases, the rank increases and saturates at the maximum value n. Numerical experiments accompanying this invention have confirmed that an exact solution can be obtained when the rank takes the maximum value n. On the other hand, when the rank is less than n, an error occurs in the predicted power density. In the numerical experiments shown later in this invention, the results shown in Table 2 confirm that when the rank of the pseudo-inverse matrix is ​​obtained to be 0.8n or greater, it is possible to reproduce the power density distribution with sufficiently acceptable accuracy.

[0053] Therefore, the number of measurement points should be at least equal to 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 necessary to further divide the inside of the fuel assembly for burnup management, the number of fuel elements will increase even more. Taking the HTTR, a high-temperature gas reactor test and research reactor, as an example, it consists of 150 hexagonal prism-shaped fuel assemblies with a distance of 36 cm between faces and a height of 58 cm. Even if we measure at least 150 detector signals, it is not practical to measure them with individual detectors, and it is necessary to move one or more detectors to measure the detector signals.

[0054] Therefore, this method assumes the measurement of detector signals from external instrumentation, internal instrumentation, and, if necessary, a combination of both.

[0055] When measuring the power distribution using only external detectors, the most ideal method for using this technique is to detect leaked neutrons from inside the reactor from outside the pressure vessel, as shown in Figure 4, while driving the detector along a helical orbit 10 to obtain detector signals at different heights, as shown in Figure 5. With this method, measurement is possible with only one detector in a series of movements, and this is a driving method also used in general X-ray CT. However, from the standpoint of manufacturability and maintainability, it is simpler to arrange multiple detectors on the circumference outside the pressure vessel and drive each detector up and down. This makes it possible to measure from virtually countless measurement points.

[0056] When used in in-reactor instrumentation, in the case of high-temperature gas reactors, the in-reactor environment is extremely high, and in the case of fast reactors, irradiation damage to the detector is significant. Therefore, it is more practical to insert the detector using a control rod guide tube when in use and store it outside the reactor when not in use, rather than keeping it permanently located in the core. In both high-temperature gas reactors and fast reactors, the control rod guide tube exists independently of the fuel, providing spatial leeway for the control rod guide tube, and allowing for design modifications to drive the detector.

[0057] Figure 6 shows an example of a high-temperature gas reactor. It assumes that the detectors are inserted from the standpipe 20 that houses the control rods. When not in use, the detectors are stored inside the standpipe 20. They can be stored at a temperature of about 300°C by the cooling flow, and irradiation damage can be prevented. When in use, they are inserted continuously to ensure a sufficient number of measurement points. In the case of an HTTR, there are 30 fuel bodies arranged radially, and there are 16 control rod guide blocks. If 16 detectors are loaded, when evaluating the power distribution on a fuel block basis, if measurement points are set at two locations in the height direction within a fuel block height of 58 cm, and every 29 cm, approximately the same number of measurement points as the fuel block can be secured. If the high-temperature region is avoided as shown in Figure 4, the number of measurement points will decrease, and a decrease in the accuracy of power distribution evaluation is expected. If this measurement accuracy is unacceptable, it is possible to avoid the decrease in accuracy by taking measures such as increasing the number of in-core detectors through design changes.

[0058] Furthermore, since the resolution of the measured output distribution can be increased by increasing the number of measurement points, using both external and internal detectors in combination enables more accurate measurement of the output distribution.

[0059] In the process of conceiving this invention, we assumed a reactor core with a longer neutron range than that of a light water reactor. However, by applying this technology to the current in-core instrumentation of light water reactors, we can expect to increase the resolution of power distribution measurements. 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 assemblies around the detector. However, considering that nearby detectors also measure leaked neutrons from the same fuel assemblies, it becomes possible to directly determine the power distribution for each assembly by performing an inverse analysis using the detector sensitivity distribution from the prior evaluation. [Industrial applicability]

[0060] Current light water reactor operators directly monitor burnup and manage in-core fuel by directly measuring power distribution using in-core instrumentation. On the other hand, high-temperature gas reactors and fast reactors have not had in-core instrumentation development or direct in-core power distribution measurement technology developed due to the high-temperature environment. This invention makes it possible for the first time to measure in-core power distribution in both types of reactors, enabling in-core fuel management similar to that of light water reactors and significantly improving safety.

[0061] In a compact reactor core system, if sufficient detector sensitivity can be ensured even to the fuel assembly in the center of the core, power distribution measurement becomes possible using only external instrumentation. By driving external detectors, the number of detectors required to obtain the necessary measurement points is reduced, and the effort required to correct for detection efficiency with a multi-detector configuration is avoided.

[0062] In reactor cores where in-core instrumentation can be installed, it becomes possible to measure power distribution over a wide range of positions within the reactor, even when the installation location is limited due to structural reasons or temperature / irradiation environment limitations. Regarding in-core instrumentation, making the in-core detectors movable not only increases the number of measurement points but also avoids degradation when the detectors are not in use. In high-temperature gas reactors and fast reactors, the control rod guide tubes are independent of the fuel assemblies, providing ample room to install driven detectors.

[0063] The performance of the present invention was applied to the High Temperature Gas Coal Reactor (HTTR) system and confirmed by simulation. For the external instrumentation, measurements were taken at a total of 180 points: 36 points on the outer circumference of the pressure vessel and 5 points at different heights. The results of measuring the power distribution of 150 fuel blocks are shown in Table 1. TIFF0007894618000026.tif47170

[0064] The average measurement error across all fuel blocks is 1.2 x 10⁻⁶. -9 %, 1.1x10 in the block showing the maximum error. -8The error was negligible, around % and a perfect match, which can be considered an exact solution, was confirmed. For the in-core instrumentation, measurements were taken at 320 points (16 detector locations and 20 height-position points). The results are shown in Table 1. The average measurement error across all fuel blocks was 3.2 x 10⁻⁶. -11 %, 2.4x10 in the block showing the maximum error. -10 A solution that can be considered an exact solution has been obtained, with an accuracy of approximately %. Table 2 shows the results when partial insertion is performed to avoid inserting the detector into the lower part of the reactor core where the temperature is high. TIFF0007894618000027.tif48170

[0065] The measurement points are fixed in position, and partial insertion is used to exclude measurement points in the lower part of the core from the 20 height-direction positions. Even with 50% insertion, a good agreement is shown, which can be considered an exact solution. With 45% and 40% partial insertion, significant errors occur, with core-average errors of 0.28% and 1.1%, and maximum local errors of 5.0% and 6.5%, respectively. As is clear from the decrease in the rank of the inverse matrix, this is an error due to insufficient measurement points rather than the position of the partial insertion, and there is room for improvement to eliminate errors even with partial insertion by arranging measurement points more densely. In this way, the power distribution of the entire core can be measured by inserting detectors into about half of the reactor, and it is possible to avoid inserting detectors into high-temperature regions. With 50% partial insertion, a decrease in the ambient temperature of the detector by about 200°C can be expected, which can greatly reduce the load on the detector from the perspective of the detector's heat resistance.

[0066] The applicability of power distribution measurement methods using external instrumentation is determined by the relationship between neutron range and core size. Table 3 shows the average neutron range for each reactor type, representing the distance traveled from neutron generation to annihilation. TIFF0007894618000028.tif21170

[0067] The diffusion distance was evaluated and compared relative to the fuel section. While it was about 7 cm for a light water reactor, it was about four times that for a high-temperature gas reactor and about six times that for a fast reactor, making it possible to estimate the power distribution of a wide range of fuel assemblies from the measured neutron information. The detector sensitivity for a light water reactor is about 40-60 cm, as shown in Figure 1. Even the design by NuScale Power, LLC in the United States, which is developing a small reactor (SMR), has a core radius of about 120 cm, so it cannot be applied to light water reactors. For high-temperature gas reactors, a wide detector sensitivity distribution can be obtained, as shown in Figure 2. Assuming that an effective core radius of approximately 130 cm can be observed from the outer periphery of the core in an HTTR, the high-power configuration of a high-temperature gas-cooled reactor (HTTR) can be addressed by increasing power density and core length. Furthermore, high-power cores adopt an annular core shape without fuel in the center for safety reasons during depressurization accidents. Therefore, this method is applicable not only to the 30 MW HTTR but also to the 50 MW, 165 MW, and 600 MW cores designed by the Japan Atomic Energy Agency (JAEA), and is expected to be applicable to most high-temperature gas-cooled reactor designs classified as SMRs (Small Modular Reactors). For fast reactors, a wider range of detector sensitivity can be expected due to the longer diffusion distance. Even if a fuel width of approximately 130 cm, similar to that of a high-temperature gas-cooled reactor, can be observed, it is still fully applicable to the PRISM (Power Reactor Innovative Small Module) reactor, a representative design of fast reactor SMRs, as its core radius is approximately 130 cm. On the other hand, in large fast reactors with a core radius of around 300 cm, it is considered difficult to apply external detectors alone.

[0068] Regarding the method using in-core instrumentation, it is not only applicable to high-temperature gas reactors and fast reactors, which are the subject of the invention, but by applying the inverse analysis method, which is the core of the inventive technology, to the data processing of the current in-core instrumentation of current light water reactors, it is possible to improve the resolution of the in-core power distribution. This is a self-evident result because the current method is an integral approach that uses the measured values ​​of the in-core instrumentation as the average value of the surrounding fuel assembly power, whereas this analysis is a differential approach that assigns and reconstructs the measured signals to each region.

[0069] In the above explanation, a pseudo-inverse matrix is ​​used to demonstrate the essential mathematical structure, and the rank of the pseudo-inverse matrix is ​​mentioned as an indicator for measuring its linearly independent components. However, even when using the least squares method, there are numerical solutions that do not use the inverse matrix, and alternative approaches to the least squares method, such as the maximum likelihood method, and multiple implementations of the inverse method exist. [Explanation of Symbols]

[0070] 10. Helical orbit 20. Standpipe

Claims

[Claim 1] A method for measuring the power distribution of a reactor core, which has a longer neutron range than a light water reactor, such as a high-temperature gas reactor or a fast reactor, based on a neutron transport equation that expresses the relationship between the power density of multiple fuel elements in the pressure vessel of a reactor, the power signals from neutron detectors at the positions of multiple neutron detectors outside the pressure vessel, and the detector sensitivity relating to the positions of the fuel elements and the neutron detectors, wherein the power distribution of the reactor core is calculated from the product of the matrix of power signals from the neutron detectors and the pseudo-inverse matrix relating to the detector sensitivity, and the power density of n multiple fuel elements i in the pressure vessel is p i The neutron detector signals from m multiple detectors at position j inside and outside the pressure vessel are then transmitted via R. j The detector sensitivity with respect to the fuel element i and the detector position j is set to w j,i Let W be the pseudo-inverse matrix for detector sensitivity. + as, The output signal from the neutron detector The pseudo-inverse matrix relating to the detector sensitivity is calculated by the above method. In a method for measuring the power distribution inside a nuclear reactor calculated by a matrix representation, A method for measuring the power distribution inside a nuclear reactor, characterized in that when the rank of the pseudo-inverse matrix is ​​0.8n (where n is the number of the multiple fuel elements), the measured value is output as a valid measurement result.

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