JPH07181283A - Calculation device and method for heavy nuclear species amount in reactor fuel - Google Patents

Abstract

PURPOSE:To calculate heavy nuclear species amount according to each origin at every moment by calculating the heavy nuclear species origin and solving the incineration equation with the use of the heavy nuclear species amount for each origin and neutron flux input in an originated species amount calculator. CONSTITUTION:The nuclear species amount according to each origin which is the original heavy nuclear species amount constituting reactor fuel and obtained with an originated species amount calculator 13, is fed-back to an originated species amount memory 11 to be stored once and input again in the device 13 in the later calculation of the heavy nuclear species calculation. The initial number density of reactor fuel is separated according to each origin and stored in an initial number density memory 10. The initial number density of reactor fuel stored in this device 10 is input in the device for each origin and input in the device 13. The changing neutron flux measured and calculated with time with a neutron flux calculator 12 is input in the device 13 for calculation and species amount according to species at every moment is calculated.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an apparatus and method for calculating the amount of heavy element nuclides in a nuclear reactor fuel, and more particularly to calculating the amount of heavy element nuclides in a nuclear reactor fuel according to its origin. The present invention relates to a device and a calculation method thereof.

[0002]

2. Description of the Related Art The regulation of nuclear materials such as nuclear fuel used in nuclear reactors is obligatory by law. Conventionally, the amount of heavy element nuclides belonging to actinide elements present in nuclear reactor fuel is controlled by the heavy element. It is managed as the total amount of nuclides.
A heavy element nuclide amount calculation device is used to calculate the heavy element nuclide amount in the reactor fuel.

A conventional heavy-element nuclide amount calculating device in a nuclear reactor fuel calculates a neutron flux distribution in the nuclear-reactor fuel, a microscopic cross-sectional area of the heavy-element nuclides constituting the nuclear-reactor fuel, and a nuclear reactor. By taking in the initial amount of heavy element nuclides that make up the fuel, the total amount of heavy element nuclides belonging to the actinide element in the reactor fuel is calculated moment by moment.

The evaluation of wear and production of heavy element nuclides in a reactor fuel is performed by spatially (three-dimensionally) dividing the reactor core into arbitrary regions, and determining the heavy element nuclides in each of the spatial division regions. This is done by calculating the quantity. The method of space division is arbitrary, but a commonly used space division method is a method of dividing the reactor core horizontally into fuel assembly units, and further dividing it into a large number of fuel segments in the vertical direction. .

Since the unit of space division utilizes the neutron flux distribution, it is general to match the division unit used when calculating the neutron flux distribution in the reactor. The conventional heavy element nuclide amount calculation device calculates the total amount of the heavy element nuclides by calculating the amount of the heavy element nuclides of the individual fuel segments thus spatially divided.

[0006] On the other hand, there are several models of production and decay of heavy element nuclides in a nuclear reactor depending on the purpose and the target core.
FIG. 3 shows a typical example of a production and decay model of heavy element nuclides used in a boiling water reactor.

The shape of the reaction of production and decay of heavy element nuclides differs depending on the nuclide, but when focusing on the amount of the heavy element nuclide which is the actinide element in the fuel segment, the amount of the nuclide can be generalized by the following differential equation. .

[0008]

[Equation 1] The average number density Ni of the nuclide i in the fuel segment of interest is represented by the equation (2).

[0009]

[Equation 2] Further, the mass Mi of the nuclide i in the fuel segment of interest can be obtained by the equation (3).

[0010]

## EQU3 ## Mi = miNiV (3) where mi is the atomic weight of the nuclide i.

A conventional heavy element nuclide amount calculating device has an initial amount of each nuclide of a nuclear reactor fuel in a fuel segment of interest, an average neutron flux of the fuel segment of interest which changes from time to time thereafter, and a fine amount for each nuclide. The above equation is solved by inputting the visual cross-sectional area, and the amount of each nuclide at every moment is calculated. This equation is called the fuel equation, and several solution methods are known (eg James J. Dud
erstadt et al., Translated by Masakuni Narita et al., "Theory and Analysis of Reactors (2)", issued by Hyundai Engineering Co., January 25, 1981, No. 58
(See page 4).

Solving the combustion equation is generally called combustion calculation. The combustion equation is nonlinear and inhomogeneous because the neutron flux and microscopic cross section are functions of time, space, and nuclide number density. Therefore, the combustion equation generally treats the effects of time dependence and space and nuclide number density dependence separately. For this treatment, the time is divided into fine steps, and for each step it is approximated that the neutron flux and microscopic cross section do not change. The width of the time step must be small enough for this approximation to hold. The differential equation (1) can be solved linearly for each time step.

By the way, generally, the measurement control of nuclear materials such as nuclear reactor fuel is the control of the location, quantity, usage status, movement, etc. of nuclear materials, and the implementation of measurement control is obligatory by law. The measurement management of nuclear materials is an essential element for the operation management of operators.

In the management of operators of nuclear power plants, it is necessary to control the amount of heavy element nuclides such as uranium and plutonium specified by law for each fuel assembly. However, once the reactor fuel is manufactured, it is impossible to measure the amount of each heavy element nuclide contained in the fuel without destroying it. Therefore, the amount of heavy element nuclides in the reactor fuel is managed by evaluating and recording by calculation, and the calculation of the amount of heavy element nuclides is performed by the conventional heavy element nuclide amount calculating device.

[0015]

By the way, most of the reactions for producing heavy element nuclides in the reactor fuel are caused by a plurality of nuclides, as is clear from the differential equation (1).
Therefore, when the amount of heavy element nuclides in the LWR fuel assembly is calculated by the conventional heavy element nuclide amount calculation device, the output amount of the nuclides is output only as a total amount regardless of the causative nuclide.

For example, uranium 2 at the time of reactor fuel production
When fuel containing 38 (U-238) burns, U-2
Some of the 38 nuclides absorb neutrons and plutonium 239
(Pu-239), but further generated Pu
-239 absorbs neutrons and plutonium 240
It changes to (Pu-240) (see FIG. 3).

On the other hand, when the fuel containing Pu-239 burns during the production of nuclear reactor fuel, part of Pu-239 absorbs neutrons and changes to Pu-240. Therefore, when the reactor fuel that shares U-238 and Pu-239 is burned during the production of the reactor fuel, Pu-240 is generated. However, in the conventional heavy element nuclide amount calculation device, the generated Pu-240 Generated P
It is impossible for u-240 to separately calculate the amount due to U-238 and the amount due to Pu-239. That is, it was not possible to calculate whether the origin nuclide of Pu-240 was U-238 or Pu-239.

When it is required to control the nuclear fuel material according to the nationality of the party to which the fuel is supplied (country of origin, enriched country, etc.) in the measurement control of nuclear fuel such as reactor fuel, 2 When a fuel containing a nuclear fuel material having more than one nationality is burned, it is necessary to calculate the amount of nuclear fuel nuclides by the nationality of the heavy element nuclide that causes them, either momentarily or at the end of burning.

The heavy element nuclide amount calculation device cannot individually calculate the amount of heavy element nuclides present in the nuclear fuel, and manages them as the total amount of each nuclide, so that fuels with two or more nationalities can be used. It was difficult to accurately calculate the amount of heavy element nuclides by their origin when burning.

The present invention has been made in consideration of the above-mentioned circumstances, and when burning a nuclear fuel, the amount of the heavy element nuclide existing in the nuclear fuel can be calculated according to its origin. An object is to provide a heavy element nuclide amount calculation device and a calculation method thereof.

Another object of the present invention is to provide an apparatus and method for calculating the amount of heavy element nuclides in a nuclear reactor fuel, which can control the amount of heavy element nuclides of a nuclear fuel material having two or more nationalities by country. To provide.

[0022]

In order to solve the above-mentioned problems, the apparatus for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to the present invention, as described in claim 1, sometimes uses the nuclear fuel in a nuclear reactor fuel. In a device for calculating the amount of heavy element nuclides in a reactor fuel that calculates the amount of heavy element nuclides belonging to the actinide element at every moment, when there are two or more sources of the heavy element nuclide in the reactor fuel, there is a fuel segment of interest. Source-specific nuclide amount storage device that stores the amount of heavy element nuclides according to origin, and source-based calculation that inputs the nuclide amount according to origin stored in this storage device and the neutron flux of the fuel segment of interest It has a nuclide amount calculation device and an output device that separates and displays the heavy element nuclide amount in the reactor fuel output from this calculation device for each source, and the nuclide amount calculation device for each source is input. Heavy element nuclides by origin And the neutron flux from solving originated by the combustion equation of heavy elements nuclides, is obtained by setting to calculate the heavy elements nuclide amount at every moment by origin.

In order to solve the above-mentioned problems, the apparatus for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to the present invention, as described in claim 2 in addition to the contents of claim 1, The reactor fuel is a mixed oxide fuel containing uranium and plutonium.
It is set so that the amount of heavy element nuclides originating from uranium and the amount of heavy element nuclides originating from plutonium are separately calculated and output during reactor fuel production, or as described in claim 3. In addition, the reactor fuel is a mixed oxide fuel containing uranium and plutonium, and the source-specific nuclide amount calculation device ignores the α decay of curium and calculates the amount of heavy element nuclides derived from uranium. It is set so as to calculate the amount of heavy element nuclides originating from plutonium, ignoring the generation of generated plutonium, and as described in claim 4, the microscopic cross-sectional area of the fuel segment of interest is tabulated. A cross-sectional area creating device for creating data is provided, and further, the method according to claim 5
As described in, the initial amount of the heavy element nuclides constituting the reactor fuel is calculated from the actual measurement value in the fuel manufacturing process and stored in the initial number density storage device.

Further, in order to solve the above-mentioned problems, the method for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to the present invention,
As set forth in claim 6, the neutron flux distribution in the reactor fuel, the microscopic cross-sections of the heavy element nuclides that make up the reactor fuel, and the heavy elements that make up the reactor fuel. This is a method that takes in the initial amount of nuclides by origin, solves the combustion equation of heavy element nuclides by origin, calculates the amount of heavy element nuclides by origin belonging to the actinide element, and outputs the calculation result.

[0025]

The apparatus and method for calculating the amount of heavy element nuclides in the reactor fuel according to the present invention solves the combustion equation as in the case of the conventional apparatus for calculating the amount of heavy element nuclides. The initial number density of the nuclear reactor fuel used is separated in advance for each origin, and the combustion equation is independently solved for each origin so that the heavy element nuclide amount at each moment can be calculated according to the origin. Accordingly, when the nuclear fuel material has two or more nationalities, it is possible to perform the measurement control of the amount of heavy element nuclide for each country.

Further, when the reactor fuel is a mixed oxide fuel (MOX fuel) containing uranium and plutonium, the initial number density of the reactor fuel given to the source-specific nuclide quantity storage device is determined by the origin of uranium and plutonium. The measurement management can be performed separately.

Furthermore, when the reactor fuel is a mixed oxide fuel containing uranium and plutonium, the reaction rate of the heavy element nuclide generation decay model is calculated from the heavy element nuclide generation decay model used to calculate the amounts of nuclides originating from each of uranium and plutonium. It is clear that even if the reaction is neglected, it does not have a significant effect on the calculation result of the combustion calculation by ignoring the reaction with a low reaction rate. It is a thing.

The apparatus for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to the present invention is provided with a cross-sectional area creating device, and the microscopic cross-sectional area required for combustion calculation in this cross-sectional area creating device is previously stored as tabular data. By preparing it, the microscopic cross-sectional area required for combustion calculation can be obtained quickly and accurately.

Furthermore, the apparatus for calculating the amount of heavy element nuclides in a reactor fuel according to the present invention uses, as the initial amount of each nuclide of the reactor fuel necessary for the combustion calculation, a value actually measured in the fuel manufacturing process. The amount of nuclide is accurately calculated.

[0030]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described below with reference to the accompanying drawings.

FIG. 1 is a block diagram showing the first embodiment of the apparatus for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to the present invention. This heavy element nuclide amount calculating device is an initial number density storage device 10 for storing an initial number density (initial atom number density) of a nuclear reactor fuel, and a nuclear reactor stored in this initial number density storage device 10. Source-specific nuclide quantity storage device 11 into which the initial number density of fuel is input according to origin, and source-specific nuclide quantity storage device 11 and neutron flux calculation device (neutron flux measurement device) 12 from moment to moment Source-based nuclide amount calculation device 13 that calculates the source-specific nuclide amount by inputting the mean neutron flux or neutron flux distribution of the source, and output that inputs and visualizes the source-specific nuclide amount calculated by the calculation device 13 And device 14. The amount of nuclides by origin is the amount of the original heavy element nuclides that make up the reactor fuel, and the amount of nuclides by origin obtained by the calculation device 13 for the amount of nuclides for origin is fed back to the storage device 11 for the amount of nuclides for origin and temporarily stored Then, in the subsequent heavy element nuclide amount calculation, it is input again to the source-specific nuclide amount calculation device 13.

The initial number density of the reactor fuel is separated according to the origin and stored in the initial number density storage device 10, and the initial number density (initial atom number density) N of the reactor fuel stored in this storage device 10 is stored. (0) indicates, as shown in step 1 (S1) of the flowchart of FIG.
After being input into 1 by origin, nuclide amount calculation device 13 by origin
Entered in. The origin-dependent nuclide amount calculation device 13 measures the neutron flux which is measured by the neutron flux calculation device 12 and is calculated as well, and is input as shown in step 2 (S2). Step 3 (S3)
From step S8, the amount of nuclides by origin is calculated every moment.

The heavy element nuclide generation and collapse model shown in FIG. 3 is used for the amount of nuclei by origin calculated by this heavy element nuclide amount calculation device.

The operation principle of Steps 3 to 8 in which the source-specific nuclide amount calculation device 13 calculates the source-specific nuclide amount every moment will be described.

To the heavy element nuclide amount calculation device 13, the nuclide amount according to origin, which is the initial number density of the reactor fuel, is separated and input from the nuclide amount storage device according to origin. The total number of origins (original heavy elements) of reactor fuel is defined as K, and the total number of heavy element nuclides treated in the combustion equation is defined as N. Then, the number density at time t of the nuclide i of origin k in the fuel segment of interest of the nuclear reactor fuel is defined as Ni, k (t). The fuel segment is a unit obtained by vertically dividing a fuel assembly mounted in the reactor core.

When the initial number density of the nuclide i of the origin k of the fuel segment of interest is Ni, k (0), the initial number density Ni (0) of all the origins of the nuclide i of the reactor fuel is

[Equation 4] Must be met. In the conventional heavy element nuclide amount calculation device, Ni (0) was input as the initial number density of the reactor fuel.

Now, the nuclear species vector of the nuclear reactor fuel is defined as follows.

[0038]

[Equation 5]

[Equation 6]

[Equation 7]

Next, the source-specific nuclide amount calculation device 13 solves the combustion equation independently for each source of the reactor fuel. Specifically, step 3 (S3) to step 6 (S6) of FIG.
Iterative calculation is performed only for the total number of sources shown in.

The combustion equation to be solved is expressed by the following equation.

[0041]

[Equation 8] Here, in the fuel segment of interest, the same cross-sectional area (σ _{j → i} ,
σ _ai ) and the same neutron flux (Φ) can be used. Now, adding equation (8), which is a combustion equation by origin for any nuclide i, for all origins,

[Equation 9]

[Equation 10] The expression (10) has exactly the same form as the expression (1). From this,

[Equation 11] It turns out that

Since the equation (11) is established for the nuclide i of an arbitrary origin, the nuclide amount vector for each origin is Nk.
For (t),

[Equation 12] Holds.

As described above, the source-specific nuclide amount calculation device 11
Can calculate the amount of nuclides by origin as shown in step 8 (S8) by solving the combustion equation by origin of equation (8), and by adding equation (12), The amount of nuclide that is not separated by origin is calculated by the source-specific nuclide amount calculation device (step 9).
(S9)).

The output device 14 is a source-specific nuclide amount calculation device 11.
Input and visualize the calculation results obtained by C
It is a monitor device such as RT.

Next, a first modified example in which the reactor fuel is applied to the MOX fuel containing uranium (U) and plutonium (Pu) is shown.

This first modification is an application of the heavy element nuclide amount calculation apparatus shown in FIGS. 1 and 2 to the measurement management of a mixed oxide fuel (MOX fuel), and the heavy element nuclide calculation apparatus shown in FIG. And the heavy element nuclide generation decay model shown in FIG. 3 is applied.

The total number of origins K of MOX fuel is 2. At the initial number density of the reactor fuel, only the uranium element nuclides are taken out, and the initial number density of uranium origin is defined as k =
Corresponds to 1. A product obtained by extracting only plutonium and americium nuclides in the initial number density is defined as an initial number density of plutonium origin and corresponds to k = 2. In the unburned MOX fuel, there is no heavy element nuclide that does not correspond to uranium (U), plutonium (Pu), or americium (Am). That is,

[Equation 13]

[Equation 14]

[Equation 15] Is.

Next, the combustion equation is solved independently for each origin. The combustion equation to be solved is as follows.

[0049]

[Equation 16] Here, at the same time, the same cross-sectional area (σ _{j → i} , σ _ai ) and the same neutron flux (Φ) are used for each origin.

Since the first modified example using MOX fuel is a special case of the first embodiment, from the equation (12), at an arbitrary time t,

[Equation 17] Holds.

As described above, the source-specific nuclide amount calculation device 13
Calculates the amount of nuclides by origin of uranium origin and plutonium origin by solving the combustion equation by origin of equation (16). Furthermore, by adding equation (17),
Calculate exactly the same amount as the amount of nuclides that are not separated by origin with conventional heavy element nuclide amount calculation devices.

The output device 14 is a source-specific nuclide amount calculation device 1.
This is to visualize the calculation result obtained in No. 3 (see FIG. 1).

The calculation results of the specific amount of the nuclides by the origin of the MOX fuel by this source-by-origin calculation device are shown in FIGS. 4 and 5.
Shown in.

MOX fuel as a nuclear reactor fuel is a fuel used for a boiling water reactor as shown in FIG. 4, and four fuel segments 20 are assembled into a nuclear reactor core as a set. It is a unit obtained by dividing the fuel assembly 21 by the required number in the vertical direction. The fuel segment 20 includes, for example, three types of fuel rods 23, 2 in a rectangular channel box 23.
4, 25 are arranged in 8 rows and 8 columns, and one water rod 26 is provided at the center.

Of the three types of fuel rods 23, 24, 25,
The fuel rod 23 is a low enriched uranium fuel rod, and the fuel rod 24 is a mixed oxide fuel rod of depleted uranium and plutonium. The remaining fuel rods 25 are low enriched uranium fuel rods containing Gd element which is a burnable poison. Sign 27 is 4
It is a control rod that is put in and taken out between a set of fuel assemblies 21.

Now, the uranium element supply party in the fuel rods 23, 24, 25 is designated as country A, and the plutonium element supply party in the fuel rods 23, 24, 25 is designated as country B. The amount of plutonium element present in the country will be controlled by the countries concerned.

FIG. 5 shows the results of calculation of changes in the average number density of plutonium isotopes in the fuel segment 20 shown in FIG.

In the conventional heavy element nuclide amount calculating device, since outputs for each nationality cannot be obtained, only the total amount of plutonium element indicated by reference numeral 31 and the amount of fissile plutonium element indicated by reference numeral 34 in FIG. 4 can be obtained. However, in the heavy element nuclide amount calculation device of the present invention, the total amount of plutonium elements 32 and 33 and the amount of fissile plutonium 35 depending on the countries involved in the supply.
And 36 are output, and it is possible to accurately measure and control the amount of heavy element nuclides for each country concerned.

Next, an example in which the calculation amount of combustion calculation of MOX fuel as a nuclear reactor fuel is reduced will be described as a second modified example of the heavy element nuclide amount calculation device.

In this second modified example, the calculation of the combustion calculation is performed from the combustion chain used for the calculation of the amounts of the nuclides originating from each of the uranium and the plutonium constituting the MOX fuel, even if the reaction is ignored because of the low reaction rate. For cases where it is clear that it does not have a significant effect on the result, in which case
By ignoring reactions with a low reaction rate, the amount of combustion calculation is reduced and simplified.

In the calculation of the amount of uranium-originating heavy element nuclides constituting the MOX fuel, the initial number density is composed of only uranium elements, so the amount of curium (Cm) elements produced is small, and therefore, due to the α decay of the curium nuclides. The amount of plutonium nuclide produced is very small. Therefore, the amount of plutonium nuclide produced by the decay of the curium nuclide is sufficiently smaller than the amount of plutonium nuclide produced by the uranium nuclide without undergoing the plutonium decay of the curium nuclide and can be ignored.

Therefore, in the calculation of the amount of heavy element nuclides originating from uranium, even if the α decay of the curium nuclide is ignored, no significant difference appears in the calculation result. However, as the number of reactions to handle is reduced, the amount of computation required is reduced. Fig. 6 shows the model of decay of heavy element nuclide generation used in the calculation of uranium-derived combustion.

On the other hand, in the calculation of the amount of heavy element nuclides originating from plutonium in MOX fuel, since the initial number density is mainly composed of plutonium element, neptunium (Np)
Since the amount of elements produced is small, the amount of plutonium nuclide produced by β-decay of the neptunium nuclide is sufficiently smaller than the amount of plutonium existing from the early stage of combustion and can be ignored.

Therefore, in the calculation of the amount of heavy element nuclides derived from plutonium, even if the generation of plutonium generated from neptunium nuclides is ignored, no significant difference appears in the calculation results. However, since the number of reactions to handle decreases,
The amount of calculation required is reduced. Fig. 7 shows the model of the decay of heavy element nuclides used in the calculation of plutonium-derived combustion.

Then, the second modified example M of the second embodiment is used.
Similar to the case of OX fuel, if the measurement is managed and calculated,
The calculation result shown in FIG. 5 was obtained. Although this calculation result is the same value as that obtained in the second modification, both the required weighing and calculation result can be reduced by about 10%.

FIG. 8 shows another embodiment of the apparatus for calculating the amount of heavy element nuclides of nuclear reactor fuel.

The heavy element nuclide amount calculating apparatus shown in this embodiment is the heavy element nuclide amount calculating apparatus shown in FIG. The other configurations are not different from those shown in the first embodiment, and therefore the same reference numerals are given and the description thereof is omitted. A cross-sectional area creation device 16 is input with a table of microscopic cross-sectional areas required for fuel calculation of a nuclear reactor fuel as a function of “burnup” and “hysteretic moderator density”.

Generally, when designing a nuclear reactor core, the average nuclear density of the fuel assemblies is determined at the time of designing the fuel assemblies, and the combustion calculation is performed for each fuel assembly using this nuclear species density. The microscopic cross-sectional area obtained by the combustion calculation for each fuel assembly is calculated as "burnup" and "hysteretic moderator density".
It is adapted to be tabulated by the cross-sectional area creating device 16 as a function of.

The cross-sectional area creating device 16 fits this tabulated microscopic cross-sectional area with the burnup and the history moderator density in the actual fuel segment, and inputs it to the source-specific nuclide amount calculating device 13. The microscopic cross-sectional area is almost constant during a slight increase in burnup, and changes linearly with changes in the hysteresis moderator density, so the error in the microscopic cross-sectional area caused by fitting is small. is there.

The initial amount of the heavy element nuclide of the nuclear reactor fuel used in this example is especially uranium 235 (U-2
35), uranium 238 (U-238), plutonium 238 (Pu-238), plutonium 239 (Pu
-239), plutonium 240 (Pu-240), plutonium 241 (Pu-241), plutonium 24
2 (Pu-242), Americium 241 (Am-24
For 1), the measured value measured in the fuel manufacturing process is used. For heavy element nuclides, which are other actinide elements, the design values determined at the time of design are used.

The microscopic cross-sectional area used for the combustion calculation is created by using the initial amount determined at the time of design, but the difference in the amount of nuclide between the actual measurement value and the design value is small. The difference between the microscopic cross-sectional areas is negligible.

The initial amounts of heavy element nuclides are measured values for uranium isotopes, plutonium isotopes, and americium 241, and for other actinide element nuclides, the design values at the time of design are used. By fitting the microscopic cross-sectional area of the reactor fuel with the burnup and the history moderator density and inputting it to the source-specific nuclide amount calculation device 13 for each source to perform the measurement control calculation, the fuel segment of interest in reality Although it is not necessary to prepare these data for each fuel segment, even though the microscopic cross section and the initial amount of nuclides differ for each, it is not necessary to prepare these data for each fuel segment. It is possible to save a lot of.

[0073]

As described above, according to the apparatus and method for calculating heavy element nuclides in a reactor fuel according to the present invention, the amount of heavy element nuclides is stored in the source-specific nuclide amount storage device according to the source. The source heavy element nuclide amount and the neutron flux of the fuel segment of interest are input and calculated by the source heavy element nuclide amount calculation device. By solving the combustion equation for each source of element nuclides and calculating the amount of heavy element nuclides at each moment by source, the amount of heavy element nuclides is accurately calculated for each source of reactor fuel, and the amount is controlled. be able to. When the nuclear fuel material of a nuclear reactor fuel has two or more nationalities, the heavy element nuclide amount can be calculated according to the origin to perform the metrological control for each country.

[Brief description of drawings]

FIG. 1 is a block diagram showing an embodiment of a heavy element nuclide amount calculating device in a reactor fuel according to the present invention.

FIG. 2 is a processing diagram showing a flowchart for calculating the amount of heavy element nuclides of a nuclear reactor fuel by origin by the heavy element nuclide amount calculation device of the present invention.

FIG. 3 is a diagram showing an example of a heavy element nuclide generation and decay model used in the heavy element nuclide amount calculation device of the present invention.

FIG. 4 is a cross-sectional view of a fuel segment mounted on the core of a boiling water reactor.

FIG. 5 is a diagram showing an output example of the atomic number densities of plutonium element and fissile plutonium element for each of the supply countries when a mixed oxide fuel is used as the reactor fuel.

FIG. 6 is a diagram showing an example of a heavy element nuclide generation and decay model for uranium origin combustion calculation when a mixed oxide fuel is used as a reactor fuel.

FIG. 7 is a diagram showing an example of a heavy element nuclide generation and decay model for plutonium-derived combustion calculation when a mixed oxide fuel is used as a reactor fuel.

FIG. 8 is a block diagram showing another embodiment of the apparatus for calculating the amount of heavy element nuclides of nuclear reactor fuel according to the present invention.

[Explanation of symbols]

10 initial number density storage device 11 nuclide amount storage device by origin 12 neutron flux calculation device 13 nuclide amount calculation device by origin 14 output device 16 cross-sectional area preparation device 20 fuel segment 21 fuel assembly 22 channel box 23 low enriched uranium rod 24 deterioration Mixed oxide fuel rod of uranium and plutonium 25 Low concentration uranium fuel rod containing Gd element which is a burnable poison 26 Water rod 27 Control rod 31 Atomic number density of plutonium element 32 Atomic number density of plutonium element originating from country B at the time of supply 33 Atomic number density of plutonium element from country A at the time of supply 34 Atomic number density of fissile plutonium element at the time of supply 35 Atomic number density of fissile plutonium element from country B at the time of supply 36 Atoms of fissile plutonium element from country A at the time of supply Number density

Claims (6)

[Claims] 1. An apparatus for calculating the amount of heavy element nuclides belonging to an actinide element in a reactor fuel, which calculates the amount of heavy element nuclides in a reactor fuel, wherein the source of the heavy element nuclide is 2 When there are two or more, the source-specific nuclide amount storage device that stores the amount of heavy element nuclides by source at a certain point in the target fuel segment, the source-specific nuclide amount stored in this storage device, and the fuel segment of interest A source-dependent nuclide amount calculation device that inputs and calculates the neutron flux of the neutron flux, and an output device that separates and displays the heavy element nuclide amount in the reactor fuel output from this calculation device for each source, Source-specific nuclide amount calculation device was set to solve the combustion equation for each source of heavy element nuclide from the input source-specific heavy element nuclide amount and the neutron flux, and calculate the heavy element nuclide amount at each moment by source. This And a device for calculating the amount of heavy element nuclides in a reactor fuel. 2. The nuclear reactor fuel is a mixed oxide fuel containing uranium and plutonium, and the source-specific nuclide amount calculation device uses plutonium and the amount of heavy element nuclides originating from uranium during the production of the reactor fuel. The amount calculation apparatus for heavy element nuclides in a nuclear reactor fuel according to claim 1, which is set to calculate and output the amount of heavy element nuclides separately. 3. The reactor fuel is a mixed oxide fuel containing uranium and plutonium, and the source-specific nuclide amount calculation device calculates the amount of heavy element nuclides derived from uranium, ignoring the α decay of curium. The apparatus for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to claim 1 or 2, wherein the amount of heavy element nuclides derived from plutonium is calculated by ignoring generation of plutonium generated from neptunium. 4. A heavy element nuclide amount calculating device in a nuclear reactor fuel according to claim 1, further comprising a cross-sectional area creating device for creating a microscopic cross-sectional area of the fuel segment of interest based on tabulated data. 5. The heavy element in the reactor fuel according to claim 1, wherein an initial amount of the heavy element nuclide constituting the reactor fuel is calculated from an actual measurement value in a fuel manufacturing process and stored in an initial number density storage device. Nuclide amount calculation device. 6. The neutron flux distribution in the reactor fuel, the microscopic cross-section of the heavy element nuclides constituting the reactor fuel, and the origin of the heavy element nuclides constituting the reactor fuel. Importing the initial amount, solving the combustion equation of heavy element nuclides by origin, calculating the amount of heavy element nuclides by origin belonging to the actinide element, and displaying the calculation result as output. Calculation method of amount of element nuclide.