JP3268099B2 - Apparatus and method for calculating heavy element nuclide amount in reactor fuel - Google Patents

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 nuclear fuel, and more particularly to the calculation of the amount of heavy element nuclides in the nuclear fuel by calculating the amount of heavy element nuclides by their origin. The present invention relates to an apparatus and a calculation method thereof.

[0002]

2. Description of the Related Art The measurement and management of nuclear materials such as nuclear fuel used in nuclear reactors are obligated by laws and regulations. Conventionally, the amount of heavy element nuclides belonging to the actinoid elements present in nuclear fuel is measured by heavy elements. 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.

[0003] A conventional apparatus for calculating the amount of heavy element nuclides in a reactor fuel includes a neutron flux distribution in the reactor fuel, a momentary microscopic cross section of heavy element nuclides constituting the reactor fuel, The initial amount of heavy element nuclides constituting the fuel is taken in, and the total amount of heavy element nuclides belonging to the actinoid element in the reactor fuel is calculated every moment.

[0004] Evaluation of the depletion and production of heavy element nuclides in the 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 space 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 into fuel assemblies in the horizontal direction, and further dividing the reactor core into a number of fuel segments in the vertical direction. .

[0005] Since the unit of the space division uses 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 amount of heavy element nuclides in each of the space-divided fuel segments to calculate the total amount of heavy element nuclides.

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

[0007] The form of the reaction of heavy element nuclide production and destruction varies depending on the nuclide, but when attention is paid to the amount of the heavy element nuclide which is an actinoid 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 expressed by 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]

[Number 3] _{Mi = mi · Ni · V /} N A ...... (3) here, mi is the atomic weight of nuclide i, _{N A} is Avogadro's number.

The conventional heavy element nuclide amount calculation apparatus includes an initial amount of each nuclide of the reactor fuel in the fuel segment of interest, an average neutron flux of the fuel segment of interest and a fine nucleus for each nuclide, which change every moment. The above equation is solved by inputting the visual cross-sectional area, and the amount of each nuclide is calculated every moment. This equation is called the combustion equation, and several solutions are known (for example, James
J. See Duderstadt et al., Reactor Theory and Analysis (2), translated by Masakuni Narita et al., Published by Hyundai Kogakusha on January 25, 1981, page 584.

Solving the combustion equation is generally called combustion calculation. The combustion equation is nonlinear and non-homogeneous because the neutron flux and microscopic cross section are functions of time, space, and nuclide number density. Therefore, the combustion equation generally deals with the effects of time dependence and the effects of space and nuclide number densities separately. For this treatment, the time is divided into small 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 be valid. The differential equation (1) can be solved linearly for each time step.

By the way, generally, the measurement and management of nuclear materials such as nuclear fuel is management of the location, quantity, use status, movement, etc. of nuclear materials, and the implementation of measurement and management is obligated by law. Nuclear material measurement and management is an essential element for the operation and management of operators.

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

[0015]

Most of the production reactions of heavy element nuclides in nuclear fuel are caused by a plurality of nuclides, as is apparent from the differential equation (1).
Therefore, when the amount of heavy element nuclides in the light water reactor fuel assembly is calculated by the conventional heavy element nuclide amount calculation device, the amount of nuclides that are output is output only as a total amount regardless of the cause nuclide.

For example, when producing reactor fuel, uranium 2
38 (U-238) when the fuel containing it burns
Some of the 38 nuclides absorb neutrons and plutonium 239
(Pu-239), but further generated Pu
-239 absorbs neutrons and makes plutonium 240
(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 sharing U-238 and Pu-239 burns during the production of the reactor fuel, Pu-240 is generated. However, in the conventional heavy element nuclide amount calculation device, Pu-240 is generated. By simply calculating the quantity, the generated P
It is impossible for u-240 to be calculated separately for the amount due to U-238 and the amount due to Pu-239. That is, it was not possible to calculate whether the source nuclide of Pu-240 was U-238 or Pu-239.

In the measurement and management of nuclear fuel such as nuclear reactor fuel, if it is required to manage nuclear fuel materials by the nationality of the country to which the fuel is supplied (country of origin, enrichment country, etc.), it is necessary to add When burning fuel containing nuclear fuel material with more than one nationality, it is necessary to calculate the amount of nuclear fuel nuclides momentarily or at the end of combustion for each nationality of the heavy element nuclide that causes it.

The heavy element nuclide amount calculation device cannot calculate the amount of the heavy element nuclides present in the nuclear fuel individually, and manages the total amount of each nuclide, so that the fuel having two or more nationalities is used. It was difficult to accurately calculate the amount of heavy element nuclides for each source when burning.

The present invention has been made in view of the above circumstances, and when nuclear fuel is burned, the amount of heavy element nuclides present in the nuclear fuel can be calculated for each source. It is an object of the present invention 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 a method for calculating the amount of heavy element nuclides in nuclear reactor fuel capable of measuring and managing the amount of heavy element nuclides of nuclear fuel materials having two or more nationalities by country. To provide.

[0022]

According to a first aspect of the present invention, there is provided an apparatus for calculating the amount of heavy nuclides in a reactor fuel according to the present invention. Reactor which spatially divides a reactor fuel into a plurality of fuel segments and calculates the amount of a heavy element nuclide belonging to an actinoid element in the reactor fuel every moment for a fuel segment of interest among the plurality of fuel segments In the apparatus for calculating the amount of heavy element nuclides in fuel, when there are two or more sources of heavy element nuclides in the reactor fuel, the nuclides by origin storing the amounts of heavy element nuclides by the source at a certain point in time of the fuel segment of interest A quantity storage device, a source-specific nuclide amount calculation device that calculates by inputting the source-specific nuclide amount stored in the storage device and the neutron flux of the fuel segment of interest every moment, An output device and for displaying to separate the heavy elements nuclides of nuclear reactor fuel et output every origin,
The source-based nuclide amount calculation device is set so as to solve a combustion equation for each source of a heavy element nuclide from the input source-based nuclide amount and the neutron flux and calculate the heavy element nuclide amount at every moment according to the source. It is.

Further, in order to solve the above-mentioned problems, a heavy element nuclide amount calculating apparatus in a nuclear reactor fuel according to the present invention has the following features. The reactor fuel is a mixed oxide fuel containing uranium and plutonium, and the source-specific nuclide amount calculation device classifies the amount of heavy element nuclides originating from uranium and the amount of heavy element nuclides originating from plutonium. Calculated separately,
And the reactor fuel is a mixed oxide fuel containing uranium and plutonium. In another embodiment, the amount of heavy element nuclide derived from uranium is calculated ignoring the decay, while the amount of heavy element nuclide derived from plutonium is calculated ignoring the production of plutonium generated from neptunium. As described in the above, a cross-sectional area creation device for creating a microscopic cross-sectional area in the fuel segment of interest based on tabulated data and inputting it to the source-specific nuclide amount calculation device is provided, and furthermore, according to claim 5, As described above, it has an initial number density storage device for storing an initial amount of the heavy element nuclide constituting the reactor fuel, and the initial number density storage device In which the period amount is input to the origin by nuclide amount storage device.

Further, the method for calculating the amount of heavy element nuclides in a nuclear reactor fuel according to the present invention is intended to solve the above-mentioned problems.
As described in claim 6, the momentary neutron flux distribution in the reactor fuel, the momentary microscopic cross-sectional area of heavy element nuclides constituting the reactor fuel, and the reactor fuel are constituted. Incorporating the initial amount of heavy element nuclides by source, solving the combustion equation of heavy element nuclides in the reactor fuel by source, calculating the amount of heavy element nuclides by the source in the reactor fuel every moment, This is a method of outputting and displaying the calculation result.

[0025]

The apparatus and method for calculating the amount of heavy element nuclides in a reactor fuel according to the present invention solves the combustion equation in the same manner as the conventional heavy element nuclide amount calculation apparatus. The initial number density of the reactor fuel to be used is separated for each source in advance, and the combustion equation is independently solved for each source, so that the amount of heavy element nuclides at every moment can be calculated for each source. Thereby, when the nuclear fuel material has two or more nationalities, the measurement and management of the heavy element nuclide amount can be performed for each country.

In the case where 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 nuclide amount storage device for each source is determined by the origin of the uranium and plutonium. Separate measurement control can be performed separately.

Further, when the reactor fuel is a mixed oxide fuel containing uranium and plutonium, the reaction rate of the reaction rate of the uranium and plutonium is calculated from the heavy element nuclide production and decay model used for calculating the amount of nuclides originating from uranium and plutonium. It is clear that ignoring the reaction due to low reaction does not significantly affect the calculation result of combustion calculation. Things.

The apparatus for calculating the amount of nuclides in a reactor fuel according to the present invention includes a cross-sectional area preparing device, and the microscopic cross-sectional area required for the combustion calculation by the cross-sectional area preparing device is stored as tabulated data in advance. By preparing, the microscopic cross-sectional area required for the 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 a value actually measured in a fuel production process as an initial amount of each nuclide of the reactor fuel necessary for a combustion calculation. The purpose is to accurately calculate the amount of heavy element nuclides.

[0030]

An embodiment of the present invention will be described below with reference to the accompanying drawings.

FIG. 1 is a block diagram showing a first embodiment of the apparatus for calculating the amount of nuclides in a reactor fuel according to the present invention. This heavy element nuclide amount calculation device includes an initial number density storage device 10 in which an initial number density (initial atomic number density) of nuclear fuel as a nuclear fuel is stored, and a nuclear reactor stored in the initial number density storage device 10. A source-specific nuclide quantity storage device 11 into which the initial number density of the fuel is input for each source, and a source-specific nuclide amount and a neutron flux calculation device (neutron flux measurement device) 12 from the source-specific nuclide quantity storage device 11 from time to time The average neutron flux or neutron flux distribution is input to calculate the source-specific nuclide amount at every moment, and the source-specific nuclide amount calculated by the source-specific nuclide amount calculator 13 is input. And an output device 14 for visualization. The source-specific nuclide amount is the original heavy element nuclide amount constituting the reactor fuel, and the source-specific nuclide amount obtained by the source-specific nuclide amount calculator 13 is fed back to the source-specific nuclide amount storage device 11 and temporarily stored. After that, in the subsequent calculation of the heavy element nuclide amount, it is input to the source-specific nuclide amount calculation device 13 again.

The initial number density of the reactor fuel is separated for each source and stored in the initial number density storage device 10, and the initial number density (initial atomic number density) N of the reactor fuel stored in the storage device 10 (0) indicates the source-specific nuclide amount storage device 1 as shown in step 1 (S1) of the flowchart of FIG.
After being input to 1 for each source, the nuclide amount calculation device 13 for each source 13
Is input to The calculated neutron flux which is measured and changed every moment by the neutron flux calculator 12 is also input to the source-based radionuclide calculator 13 as shown in step 2 (S2). Step 3 (S3)
, The calculation of step 8 (S8) is performed, and the source-specific nuclide amount at every moment is calculated.

The heavy element nuclide production decay model shown in FIG. 3 is used for the source-specific nuclide amount calculated by the heavy element nuclide amount calculation apparatus.

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

The source-specific nuclide quantity storage unit 11 receives the source-specific nuclide quantity, which is the initial number density of the reactor fuel, separated into the heavy element nuclide quantity calculation unit 13 for each source. The total number of reactor fuel sources (original heavy elements) 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 of the nuclide i of the origin k in the fuel segment of interest in the reactor fuel at the time t is defined as Ni, k (t). The fuel segment is a unit obtained by dividing a fuel assembly mounted on a reactor core in a vertical direction.

Assuming that 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 the total origin of the nuclide i of the reactor fuel is:

(Equation 4) Must be satisfied. In the conventional heavy element nuclide amount calculation apparatus, Ni (0) is input as the initial number density of the reactor fuel.

Now, the nuclide quantity vector of the reactor fuel is defined as follows.

[0038]

(Equation 5)

(Equation 6)

(Equation 7)

Next, the source-specific nuclide amount calculator 13 solves the combustion equation independently for each source of the reactor fuel. Specifically, from step 3 (S3) to step 6 (S6) in FIG.
Is repeated for the total number of origins.

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

[0041]

(Equation 8) Here, within the fuel segment of interest, the same cross-sectional area (σ _{j → i} ,
σ _ai ), the same neutron flux (Φ) can be used. Now, when equation (8), which is a combustion equation for each source for an arbitrary nuclide i, is added for all sources,

(Equation 9)

(Equation 10) Equation (10) has exactly the same form as equation (1). From this,

[Equation 11] It turns out that it is.

Since the formula (11) holds for a nuclide i of an arbitrary origin, the nuclide quantity vector for each source at every moment is Nk
About (t),

(Equation 12) Holds.

As described above, the nuclide amount calculation device 11 for each source
Can calculate the source-specific nuclide amount as shown in step 8 (S8) by solving the source-specific combustion equation of equation (8). Further, by adding equation (12), The amount of nuclide which is exactly the same as the amount of nuclide which is not separated by source is calculated by the source-specific nuclide amount calculator (step 9).
(S9)).

The output unit 14 is a unit for calculating the nuclide amount by source 11
C to input and display the calculation result obtained by
A monitor device such as an RT.

Next, a first modification in which the reactor fuel is applied to MOX fuel containing uranium (U) and plutonium (Pu) will be described.

In this first modification, the heavy element nuclide amount calculation apparatus shown in FIGS. 1 and 2 is applied to the measurement control of mixed oxide fuel (MOX fuel), and the heavy element nuclide calculation apparatus shown in FIG. And the heavy element nuclide production decay model shown in FIG. 3 is applied.

The total origin number K of the MOX fuel is 2. From the initial number density of the reactor fuel, only the uranium element nuclide was taken out and defined as the initial number density of uranium origin.
Corresponds to 1. A product obtained by taking out only plutonium and americium nuclides in the initial number density is defined as an initial number density originating from plutonium, and corresponds to k = 2. There is no heavy element nuclide that does not correspond to any of uranium (U), plutonium (Pu), and americium (Am) in the unburned MOX fuel. That is,

(Equation 13)

[Equation 14]

(Equation 15) It is.

Next, the combustion equation is independently solved 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 the MOX fuel is a special case of the first embodiment, at any time t from the equation (12),

[Equation 17] Holds.

As described above, the nuclide amount calculating device 13 for each source 13
Calculates the source-specific nuclide amounts of uranium and plutonium by solving the source-specific combustion equation of equation (16). Further, by performing the addition of equation (17),
The same amount as the amount of nuclide that is not separated for each source by the conventional heavy element nuclide amount calculator is calculated.

The output device 14 is a device for calculating the amount of nuclides by source 1
3 to visualize the calculation results obtained (see FIG. 1).

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

As shown in FIG. 4, MOX fuel as a reactor fuel is a fuel used for a boiling water reactor, and four fuel segments 20 are mounted in a reactor core in groups of four. This is one unit obtained by dividing the required number of fuel assemblies 21 along the vertical direction. For example, three types of fuel rods 23, 2 are placed in a rectangular cylindrical 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. Symbol 27 is 4
These are control rods that are put in and taken out between a set of fuel assemblies 21.

Now, the country that supplies the uranium element in the fuel rods 23, 24, and 25 is Country A, the country that supplies the plutonium element in the fuel rods 23, 24, and 25 is Country B, and the fuel segment 20 The amount of plutonium elements present in each country will be managed by the supplying country.

FIG. 5 shows the result of calculation of the change in the average number density of the plutonium isotope in the fuel segment 20 shown in FIG.

In the conventional heavy element nuclide amount calculation apparatus, since the output for each nationality cannot be obtained, only the total amount of plutonium element indicated by reference numeral 31 in FIG. 5 and the amount of fissile plutonium element indicated by reference numeral 34 can be obtained. However, in the heavy element nuclide amount calculation apparatus of the present invention, the total amount of plutonium elements 32 and 33 and the amount of fissile plutonium 35
And 36 are output, and the amount of heavy element nuclides can be accurately measured and controlled for each supplying country.

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

This second modification is based on the calculation of combustion calculation from the combustion chain used for calculating the respective nuclide amounts of uranium and plutonium constituting the MOX fuel, ignoring the reaction due to its low reaction rate. In cases where it is clear that the result will not be significantly affected,
By ignoring the reaction having a low reaction rate, the calculation amount of the combustion calculation is reduced and simplified.

In calculating the amount of uranium-derived heavy element nuclides constituting the MOX fuel, since the initial number density consists only of the uranium element, the amount of the curium (Cm) element produced is small. The amount of plutonium nuclides produced is very small. For this reason, the amount of plutonium nuclide generated by the decay of the curium nuclide is sufficiently small and negligible as compared with the amount of the plutonium nuclide generated from the uranium nuclide without undergoing the plutonium decay of the curium nuclide.

Thus, in the calculation of the amount of heavy element nuclides derived from uranium, no significant difference appears in the calculation results even if the α decay of the curium nuclide is ignored. However, as the number of reactions to be handled is reduced, the amount of computation required is reduced. FIG. 6 shows a heavy element nuclide production decay model used in the uranium origin combustion calculation.

On the other hand, in the calculation of the amount of heavy element nuclides originating from plutonium in the MOX fuel, the initial number density is mainly composed of plutonium element, so that neptunium (Np)
The amount of element produced is small, and therefore the amount of plutonium nuclide produced by β decay of neptunium nuclide is sufficiently small and negligible compared to the amount of plutonium present from the beginning of combustion.

From the above, in the calculation of the amount of heavy element nuclides originating from plutonium, even if the production of plutonium generated from neptunium nuclides is ignored, no significant difference appears in the calculation results. However, since the number of reactions to be handled decreases,
The amount of computation required is reduced. FIG. 7 shows a heavy element nuclide production decay model used in the calculation of plutonium origin combustion.

Then, this second modified example is referred to as M of the second embodiment.
When the calculation is performed by measuring and managing in the same way as in the case of OX fuel,
The calculation result shown in FIG. 5 was obtained. This calculation result has the same value as that obtained in the second modification, but both the required amount of calculation and the calculation result can be reduced by about 10%.

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

The heavy element nuclide amount calculation apparatus shown in this embodiment is obtained by adding the heavy element nuclide amount calculation apparatus shown in FIG. The other configuration is not different from that shown in the first embodiment, so the same reference numerals are given and the description is omitted. A table of the microscopic cross-sectional area required for the fuel calculation of the reactor fuel as a function of “burnup” and “history moderator density” is input to the cross-sectional area creation device 16.

In general, when designing a nuclear reactor core, the average nuclide number density of the fuel assembly is determined at the time of fuel assembly design, and the combustion calculation of each fuel assembly is performed using the nuclide number density. The microscopic cross-sectional area obtained by the combustion calculation for each fuel assembly is called "burnup" and "history moderator density".
Are tabulated by the cross-section creation device 16 as a function of

The cross-sectional area preparing device 16 fits the tabulated microscopic cross-sectional area with the burn-up in the actual fuel segment and the hysteresis moderator density, and inputs it to the source-specific nuclide amount calculating device 13. Since the microscopic cross section is almost constant during a slight increase in burnup and changes linearly with changes in hysteresis moderator density, errors in the microscopic cross section caused by fitting are small. is there.

The initial amount of the heavy element nuclide of the reactor fuel used in this embodiment is, in particular, 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)
As for 1), an actual measurement value measured in the fuel manufacturing process is used. For heavy element nuclides that are other actinoid elements, design values determined at the time of design are used.

Although the microscopic cross section used in the combustion calculation is created using the initial quantity determined at the time of design, the difference in the nuclide quantity between the actually measured value and the design value is so small that it is generated by the difference. The difference in the microscopic cross-sectional area is negligible.

Then, the initial amount of the heavy element nuclide is measured by the uranium isotope, the plutonium isotope, and the americium 241, and the other actinide element nuclides are designed by using the design values at the time of design. By fitting the microscopic cross-sectional area of the reactor fuel with the burn-up and the history moderator density and inputting it into the source-specific nuclide quantity calculation device 13 for each source and performing the measurement management calculation, the fuel segment of interest that is actually focused on Although there is no need to prepare these data for each fuel segment, despite the fact that the microscopic cross-sectional area and the initial amount of nuclides are different for each, the computational resources required for the calculation of heavy element nuclide amounts on a whole core scale Large savings are possible.

[0073]

As described above, according to the apparatus and method for calculating heavy element nuclides in a nuclear reactor fuel according to the present invention, the amount of heavy element nuclides is stored in the storage device for each nuclide amount by source. The source-specific heavy element nuclide amount and the neutron flux of the fuel segment of interest at each moment are input and calculated by the source-specific heavy element nuclide amount calculator, which calculates the heavy element nuclide amount by source and the neutron flux. By solving the combustion equation for each source of elemental nuclides and calculating the amount of heavy elementary nuclides from time to time for each source, the amount of heavy elementary nuclides is accurately calculated for each arbitrary source of reactor fuel and measured and managed. be able to. When the nuclear fuel material of the reactor fuel has two or more nationalities, the amount of heavy element nuclides can be calculated for each source, thereby enabling quantitative control for each country.

[Brief description of the drawings]

FIG. 1 is a block diagram showing one embodiment of an apparatus for calculating the amount of heavy element nuclides in a reactor fuel according to the present invention.

FIG. 2 is a processing diagram showing a flowchart for calculating a heavy element nuclide amount of a reactor fuel for each source 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 production 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 a core of a boiling water reactor.

FIG. 5 is a diagram showing an output example of the atomic number density of the plutonium element and the fissile plutonium element by supply country in the case of using a mixed oxide fuel as the reactor fuel.

FIG. 6 is a diagram showing an example of a heavy element nuclide production and decay model for uranium-originated 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 production decay model for plutonium-derived combustion calculation in the case of using a mixed oxide fuel as a reactor fuel.

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

[Explanation of symbols]

REFERENCE SIGNS LIST 10 Initial number density storage device 11 Source-specific nuclide amount storage device 12 Neutron flux calculator 13 Source-specific nuclide amount calculation device 14 Output device 16 Cross-sectional area creation 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-enriched uranium fuel rod containing Gd element, which is a burnable poison 26 Water rod 27 Control rod 31 Atomic density of plutonium element 32 Atomic density of plutonium element originating from Country B at the time of supply 33 Atomic density of plutonium element originating from Country A at the time of supply 34 Atomic density of fissile plutonium element originating from Country A at the time of supply 35 Atomic density of fissile plutonium element originating from Country B at the time of supply 36 Atoms of fissile plutonium element originating at Country A at the time of supply Number density

──────────────────────────────────────────────────続 き Continued on front page (58) Field surveyed (Int.Cl. ⁷ , DB name) G21C 17/06

Claims (6)

(57) [Claims] 1. A reactor fuel provided in a reactor core is spatially divided into a plurality of fuel segments, and the amount of a heavy element nuclide belonging to an actinoid element in the reactor fuel at each moment is determined by the plurality of fuel segments. In the heavy element nuclide amount calculation device in the reactor fuel for calculating the fuel segment of interest out of the segments, when there are two or more sources of heavy element nuclides in the reactor fuel, the source of the fuel segment of interest at a certain point in time A source-specific nuclide amount storage device that stores the heavy element nuclide amount of the source element, and a source-specific nuclide amount calculation device that calculates by inputting the source-specific nuclide amount stored in the storage device and the neutron flux of the fuel segment of interest every moment And an output device that separates and displays the heavy element nuclide amount in the reactor fuel output from the calculation device for each source, and the source-specific nuclide amount calculation device is input. The amount of heavy element nuclide in the reactor fuel is characterized in that the combustion equation is solved for each source of heavy element nuclide from the source-specific nuclide amount and the neutron flux and the heavy element nuclide amount at every moment is calculated for each source. Calculation device. 2. The nuclear reactor fuel is a mixed oxide fuel containing uranium and plutonium, and the apparatus for calculating the amount of nuclides by source comprises the amounts of heavy elements nuclides originating from uranium and heavy elements originating from plutonium. 2. The apparatus according to claim 1, wherein the apparatus calculates and outputs the nuclide amount separately from the nuclide amount. 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 uranium-derived heavy element nuclides ignoring 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 apparatus is configured to calculate the amount of plutonium-derived heavy element nuclides while ignoring the production of plutonium generated from neptunium. 4. The reactor fuel according to claim 1, further comprising a cross-sectional area creating device for creating a microscopic sectional area in the fuel segment of interest based on tabulated data and inputting the table to the source-specific nuclide amount calculation device. Heavy element nuclide amount calculation device. 5. An initial number density storage device for storing an initial amount of the heavy element nuclides constituting the reactor fuel, wherein the initial amount stored in the initial number density storage device is the source-specific nuclide amount. 2. The apparatus according to claim 1, wherein the apparatus is inputted to a storage device. 6. An instantaneous neutron flux distribution in a reactor fuel, an instantaneous microscopic cross-sectional area of a heavy element nuclide constituting the reactor fuel, and a heavy element nuclide constituting the reactor fuel. Taking in the initial amount by source, solve the combustion equation of heavy element nuclides in the reactor fuel by source, calculate the amount of heavy element nuclides in the reactor fuel by time, and output the calculation result A method for calculating the amount of heavy element nuclides in a reactor fuel, the method comprising displaying the information.