JPS61213690A - Monitor device for distribution of output from 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 a nuclear reactor power distribution monitoring device for monitoring the power distribution of a nuclear reactor.

[Technical background of the invention and its problems] In order for a nuclear reactor to maintain its health and exhibit the necessary performance, it is necessary to monitor the current state of core performance, and for this purpose, process control computers are generally used. There is. Conventional process control computers input thermal neutron flux measurements within the reactor core and current state data of the reactor core, and calculate the power distribution within the reactor core based on these inputs. Here, the thermal neutron flux measurement value is 4
It is obtained by thermal neutron flux measuring instruments arranged at intervals surrounded by the fuel assemblies (hereinafter referred to as corner gaps), and the current state data of the core is obtained by measuring the coolant temperature at the core inlet and outlet, the core inlet enthalpy, and the flow rate. It is obtained from measurements of control rod positions, etc. The power distribution of the reactor core calculated in this way is one of the main parameters required to monitor the health of the reactor.

Therefore, the power distribution within the reactor core obtained by the above-mentioned process control computer is completely dependent on the measured value of the thermal neutron flux measuring device. Therefore, if all thermal neutron measuring instruments are in normal condition, the calculated output distribution will be very close to the actual output distribution.

FIG. 2 shows a part of a conventional thermal neutron measuring instrument arranged in a reactor core. Inside the reactor, a large number of fuel assemblies (hereinafter referred to as bundles) 2 that make up the reactor core 1 are arranged in an overall cylindrical shape, and relatively rapid changes in reactivity can be controlled by control rods 3. It is about to be done. A conduit 5 is arranged at the center of the corner gap 4 surrounded by the bundle 2.

A thermal neutron flux measuring device 6 is fixed in a conduit 5 in the corner gap 4 shown in the figure at a predetermined position in the axial direction of the core 1, but other thermal neutron measuring devices (not shown) move inside the conduit 5. It is now possible to do so.

The latter thermal neutron flux measuring device can measure thermal neutron flux at any location in the axial direction. In the following description, the former measuring instrument will be referred to as a fixed measuring instrument, and the latter measuring instrument will be referred to as a mobile measuring instrument.

By the way, if the conduit 5 in the corner gap 4 is bent due to cooling water flow or the like in the nuclear reactor, these measuring instruments will be displaced in the radial direction within the corner gear 4 as a result. On the other hand, the thermal neutron flux is generally not flat within the corner gap 4, and is large in the central portion and small in the vicinity of the bundle 2. Therefore, if these measuring instruments are deflected, accurate measurements regarding thermal neutron flux cannot be obtained. When such a situation occurs, for example, data at positions that should originally give exactly the same measurement value may be input to the process control computer as mutually different data. In this case, the correct power distribution of the reactor may not be obtained, which is expected to have an unfavorable effect on the health of the fuel and the operating rate of the reactor.

[Object of the Invention] The present invention has been made in view of the above circumstances, and its purpose is to obtain the correct power distribution of a nuclear reactor even if the measuring instrument deviates from the center position of the corner gap. 'An object of the present invention is to provide a reactor power distribution monitoring device.

[Summary of the Invention] In order to achieve the above object, the present invention provides a measurement method that is set in the axial direction at the center of the interval surrounded by four fuel assemblies in a reactor core in which a large number of fuel assemblies are arranged. A gamma-ray flux measuring device that measures gamma-ray flux at a set point and a pre-built conversion formula for the gamma-ray flux measurement values obtained from this gamma-ray flux measuring device are used to measure the gamma-ray flux loaded in the four fuel assemblies. A nuclear reactor equipped with an adjacent fuel rod average power conversion device that converts the average power of a plurality of fuel rods close to a set point, and which monitors a reactor power distribution calculated based on the average power of the fuel rods. This invention relates to an output distribution monitoring device.

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

FIG. 1 shows a system diagram of a nuclear reactor power distribution monitoring device according to the present invention. As shown in the figure, inside the reactor 10 there is a fixed thermal neutron flux measuring device 13 and a mobile gamma ray flux measuring device 14.
Three types of measuring instruments are arranged: and core current data measuring instrument 15. The fixed thermal neutron flux measuring device 13 is installed in the nuclear reactor 1.
It is used to constantly monitor the output of 0. The gamma ray flux measuring device 14 uses a measuring device that utilizes the ionization phenomenon of gas due to gamma rays, and is used periodically or as necessary to determine the output distribution, and there are only a few points in the axial direction within the corner gap. Used for calibrating fixed measuring instruments.

Furthermore, the core current data measuring device 15 measures the total flow rate of the coolant;
It is used to monitor inlet and outlet temperatures, core pressure, control rod positions, etc.

By the way, the measured value 16 outputted from the thermal neutron flux measuring device 13 and the core current state data 11 outputted from the core current state data measuring device 15 are collected in a data sampler 18, and processed by later-described devices 19 to 21. After it is done,
The results are output from the input/output device 22.

Now, the gamma-ray flux measuring device 23 performs control such as applying a predetermined voltage to the gamma-ray flux measuring device 14, and from this, the output signal 2
Get 4. The output signal 24 is subjected to necessary processing using an amplifier (not shown), etc., and is input to the data sampler 18 as a measured value 24 of the γ-ray flux at each corner gap in the core 11. The measured value 24 obtained using the gamma ray flux measuring device 14 is inputted to the adjacent fuel rod average power conversion device 19 via the data sampler 18 together with the core current state data 17 obtained by the core current state data measuring device 15.

The adjacent fuel rod average power conversion device 19 uses the built-in conversion formula to calculate the input measurement value 25 and core current data 1.
1 to the adjacent fuel rod average power at each measurement set point. The conversion formula is expressed by the following formula (1).

P (i, j) −A (i, j) xT(i, j
)...(1) Here, i is the axial measurement set point number in the corner gap, j is the corner gap number, T(i, j> is the measured value at the position (i, j>), P ( i,
j) is the average output of adjacent fuel rods at the (i, j) position, A(+, j> is the coefficient for converting the measured value at the (i, j) position into the average output of adjacent fuel rods, and this implementation In this example, a total of 16 fuel rods, 4 of which are closest to the conduit 12 in each bundle, are selected as the proximal fuel rods. Accounting for over 70% of the total,
This is because it becomes possible to calculate the reactor power distribution more accurately than when the number of fuel rods is less than this.

Now, the above equation (1) has a form similar to that used in conventional reactor power monitoring equipment that uses a thermal neutron measuring device as a mobile measuring device, and the In this method, a four-bundle system is constructed with a thermal neutron flux measuring device placed in the center, and for example, the two-dimensional diffusion method is used to calculate the measured value of the thermal neutron flux measuring device at rated output and the average output of adjacent fuel rods near this measuring device. By calculating the ratio of these values, the coefficient for converting the measured value to the average output of adjacent fuel rods was determined in advance.

By the way, in the reactor power distribution monitoring device according to the present invention, in addition to calculating the average power of adjacent fuel rods near the measuring device by calculating the two-dimensional fuel assembly nuclear constant of a four-fuel assembly system for the rated power, which is currently performed, In addition, it becomes necessary to calculate in advance the measured value of the gamma ray flux measuring device at the rated output using, for example, a transport calculation method. Then, the ratio of the calculated value of the gamma-ray flux measurement value described above to the average output of the adjacent fuel rods near the gamma-ray flux measuring device obtained by two-dimensional diffusion for the four fuel assembly system is taken, and the coefficient A (i, j) is calculated as before. This will need to be calculated.

Therefore, the method for calculating the coefficient value of the gamma ray flux measuring device at the rated output is as follows.

In other words, the number of fission or absorption reactions of each nuclide in each region of each fuel rod or channel box obtained from the normal calculation of the nuclear constant of a single fuel assembly in an infinite lattice system is calculated based on the unit fission or absorption reaction of that nuclide. The gamma rays generated at each point are determined depending on the energy. Using this gamma-ray source, we performed gamma-ray transport calculations for a system in which a gamma-ray flux measuring device is placed at the center of an infinite lattice system of single fuel assemblies, and the gamma-ray flux measuring device is surrounded by a single type of fuel assembly. Calculate the measured value when

On the other hand, in the same way as when using a reactor power monitoring system using a thermal neutron flux measuring device, we performed two-dimensional diffusion calculations for a four-fuel assembly system for the fuel assembly arrangement surrounding the actual measuring device, and calculated the output power of each fuel assembly. You get an allocation. By averaging the measured values when the measuring device is placed at the center position of the single fuel assembly infinite lattice system obtained earlier, using this output distribution as a weight, the center of the four fuel assemblies at the rated output γ existing at the position
The measurement value of the flux measuring device can be calculated in advance.

This is due to the characteristic of gamma rays that the level of the gamma ray source rises and falls depending on the output of each fuel assembly, and that each gamma ray source is independent of the gamma ray flux in each region generated by it. be.

In addition, by using the results of gamma-ray transport calculations for an infinite lattice of a large number of 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 values can be calculated using simple calculations from parameters such as the average enrichment and void fraction of the fuel assembly, there is usually no need to calculate gamma-ray transport for a single fuel assembly infinite lattice system. It happens.

In addition, in the normally proposed reactor power distribution monitoring equipment that uses a gamma-ray flux measuring device, when calculating the coefficient for converting the gamma-ray flux measurement value to the average power of adjacent fuel rods, a single fuel set is used as described above. The gamma-ray flux measurements calculated in a system where the measuring device is placed at the center of the infinite grid system are averaged using the output distribution obtained by the four-fuel bundle combination two-dimensional diffusion calculation in the system where the system is actually placed as weights. Instead, it is calculated by taking a simple average. Therefore, it can be said that the conversion coefficient of the reactor power distribution monitoring device according to the present invention has higher accuracy, and the power monitoring accuracy thereof is improved.

The adjacent fuel rod average power 26 at the measurement set point determined by the fuel rod average power conversion device 19 is input to the fuel assembly average power calculation device 20 . Fuel assembly average output calculation device 2
0, the measurement set point 4 fuel assembly average power 27 is calculated and input to the measurement set point fuel assembly output calculation device 21.

The measured set point fuel assembly output calculation device 21 calculates the measured set point bundle output 28 . Note that the four-fuel bundle combined two-dimensional diffusion calculation is also performed to obtain the coefficients of the power calculation device described above, and in this respect there is no difference from the case of a reactor power distribution monitoring device that uses a thermal neutron flux measuring device. do not have. That is, the coefficients used inside the fuel assembly average power calculation device 20 and measurement set point fuel assembly power calculation device 21 used to determine the measured set point bundle output 28 from the adjacent fuel rod average power 26 are degree, fuel type,
Since it depends on the void ratio, fuel assembly arrangement, and control rod insertion pattern, two-dimensional diffusion calculation is performed in advance for combining four bundles surrounding the neutron detector and for each insertion pattern of four control rods surrounding the four bundles. was performed for burnup and void ratio at several points, and the above fuel assembly average power calculation device 2
0 and the coefficients used inside the measurement set point fuel assembly output calculation device 21 are obtained by footing the results of the four-fuel bundle combined two-dimensional diffusion calculation.

The measurement set point bundle output 28 is then output to the input/output device 2
2 is input. The output section of the input/output device 22 is equipped with a CRT and a line printer, and outputs the core power distribution.

The measurement set point bundle output 28 is also provided to a thermal limit calculation device 30 as required. Thermal limit value calculation device 30 calculates various thermal limit values for ensuring the soundness of the fuel.

On the other hand, the input/output section of the input/output device 22 outputs a calculation instruction signal 29 in order to obtain necessary data on the reactor power distribution. The calculation instruction signal 29 is output periodically, for example, once every hour, and is supplied to the data sampler 18 and the adjacent fuel rod average power conversion device 19 to start data transfer and core power distribution calculation. Of course, an operator on the reactor side can generate the calculation instruction signal 29 at a desired time to determine the reactor power distribution, if necessary.

In this way, according to the reactor power distribution monitoring device of this embodiment, as described above, the thermal neutron flux measuring device 13 and the γ-ray flux measuring device 14 are used together, and of these, the thermal neutron flux measuring device 13 is used for the atomic Several units are permanently installed in each corner gap to monitor the power distribution inside the reactor, and measurements are taken constantly. On the other hand,
The γ-ray flux measuring device 14 is employed as a mobile measuring device in this embodiment, and is used for calibrating the output distribution determined by the fixed thermal neutron flux measuring device 13. Measurement by the movable gamma-ray flux measuring instrument 14 is performed when the output distribution changes significantly, such as when a control rod pattern changes, and provides measurement values at multiple points while moving in the axial direction of each corner gap.

Therefore, a measured value of gamma-ray flux is not normally obtained, but in this regard, a practical measured value of gamma-ray flux is obtained by the following method. That is, the constant values on both sides of the fixed thermal neutron flux measuring device 13 and the mobile gamma ray flux measuring device 14 at the same measurement point do not match due to differences in measurement targets, differences in detection efficiency, and the like. However, it is possible to obtain a coefficient for each measurement point using the thermal neutron flux measuring device 13 so that the two coincide with each other. Therefore, the measured value of the gamma ray flux at each measurement setting point at each measurement point is the value obtained by multiplying the current measurement value of the thermal neutron flux measuring device 13 within that corner gap by this coefficient, and This can be determined by interpolating the difference between the measured value of the gamma-ray flux measuring device 14 at the closest point in time to the current point in time at the position where the gamma-ray flux measuring device 14 is placed, and adding it to the measured value of the gamma-ray flux measuring device 14.

In other words, even if there are different types of measuring instruments, such as a fixed measuring instrument and a mobile measuring instrument, as in the reactor power distribution monitoring system of this embodiment, the correction coefficient of the fixed measuring instrument can be calculated for the measured value of the mobile measuring instrument. Therefore, no problems will occur. Of course, in this example, the gamma-ray flux is measured using a gamma-ray flux measuring device as a moving measuring device that serves as a reference for each measurement set point, so even if the measuring device is shifted from the center of the corner gap, the measurement will still be possible. The value is accurate. Therefore, the accuracy of the reactor power distribution monitoring device will be significantly improved. In this way, sufficient accuracy can be obtained by using the γ-ray flux measuring device only as a mobile measuring device.

In addition, since the present invention uses a gamma ray flux measuring device as described above, it is necessary to convert the measured gamma ray flux value into the average output of the fuel rods near the measurement set point. However, this is based on the conventional two-dimensional diffusion calculation for four fuel assemblies, which is based on the average output of fuel rods near the set point, the power distribution of each fuel assembly, and the difference between the gamma ray source and gadolinia rod when calculating a single fuel assembly. It can be easily obtained using gamma ray flux measurements at rated output, which are related to gadolinia concentration and position, average void fraction, and burnup, and is comparable to conventional equipment.

[Effects of the Invention] As explained above, according to the reactor power distribution monitoring device of the present invention, even if the position of the measuring device in the conduit in the corner gap deviates from the center position of the corner gap, γ
By using a flux measuring device, accurate measurement values can be obtained, and a more accurate reactor power distribution can be stably obtained than when using a thermal neutron flux measuring device. Furthermore, the calculation method of the adjacent fuel rod average power conversion coefficient of the gamma ray flux measurement value becomes more accurate, and the accuracy of the reactor power distribution monitoring device is further improved compared to the conventional one, which is an excellent effect.

[Brief explanation of the drawing]

FIG. 1 is a block diagram of an embodiment of the present invention, and FIG. 2 is a perspective view of conventional thermal neutron measuring instruments arranged in a reactor core. DESCRIPTION OF SYMBOLS 10... Nuclear reactor, 11... Core 12... Corner gap 13... Fixed thermal neutral flux measuring device 14... Mobile gamma ray flux measuring device 15... Core current data measuring device 18. ... Data sampler 19 ... Adjacent fuel rod average power conversion device 20 ... Fuel assembly average power calculation device 21 ... Measurement set point fuel assembly power calculation device 22 ... Input/output device 23 ... γ-ray flux measuring device (8733) Agent: Yoshiaki Inomata, patent attorney (and 1 others)
given name)

Claims (1)

[Claims] A gamma-ray flux measuring device that measures gamma-ray flux at a measurement set point set in the axial direction at the center of the interval surrounded by four fuel assemblies in a core in which a large number of fuel assemblies are arranged, and this gamma-ray flux. Proximity fuel converts the gamma ray flux measurement value obtained from the measuring device into the average output of multiple fuel rods near the measurement set point loaded in the four fuel assemblies using a pre-built conversion formula. 1. A reactor power distribution monitoring device, comprising: a rod average power conversion device, and configured to monitor a reactor power distribution calculated based on the average power of the fuel rods.