JPH1184060A - Method and device for calculating output distribution of fast reactor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a method for calculating the output distribution of a fast reactor by which the output distribution and neutron flux distribution of the reactor can be calculated in detail by taking the heterogeneousness of a fuel assembly for irradiation into consideration and the changes of the distributions by burning can be calculated in a short time and a calculating device used for the method. SOLUTION: In a method for calculating output distribution of fast reactor, the calculation of three-dimensional neutron diffusion is performed for calculating the neutron flux distribution and output distribution of the whole core of a fast reactor by using a nuclear cross-sectional area prepared based on the composition of each fuel assembly in the core of the fast reactor. Then the calculation of two-dimensional neutron diffusion is performed for calculating the neutron flux distribution and output distribution at every two-dimensional cross section in the horizontal direction of the core by using the axial buckling of each assembly and nuclear fusion radiation source distribution of the fuel assemblies of the core obtained as a result of the three- dimensional neutron diffusion calculation by performing node division in more detail than that performed in the three-dimensional neutron diffusion calculation.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method and apparatus for calculating a power distribution of a fast reactor, and more particularly, to a fast reactor for efficiently calculating a detailed power distribution and its history in an irradiation assembly of a fast reactor. And an apparatus for calculating an output distribution.

[0002]

2. Description of the Related Art In recent years, various fuel assemblies have been developed in order to achieve high performance such as high linear output and high burn-up for core fuel of fast reactors. Then, the newly developed fuel assembly is loaded into a fast reactor core and an irradiation characteristic test is performed to confirm its performance.

When performing the irradiation characteristics test, a fast reactor core is loaded with a core fuel assembly for maintaining criticality, securing output, and continuing operation, and a fuel assembly for an irradiation test. Is done. In addition, special assemblies may be loaded for testing core materials such as higher performance absorbers and shields. Hereinafter, a special assembly including these irradiation fuels and materials is referred to as an “irradiation assembly”.

[0004] The core fuel assembly has a shape in which many identical fuel pins are bundled, and the irradiation assembly has various structures for adjusting its purpose and irradiation conditions, handling, and adjusting nuclear properties as a fuel. And generally has a more complicated structure than a core fuel assembly.

FIGS. 3A and 3B are cross-sectional views of typical core fuel assemblies and irradiation assemblies, respectively. As shown in FIG. 3 (A), the core fuel assembly is configured by accommodating a large number of core fuel pins 6 inside a trumpet tube 7.

Further, as shown in FIG. 3B, the irradiation assembly is divided into six small assemblies, that is, irradiation segments 8. Each segment 8 includes an irradiation core fuel pin 9, an MA mixed fuel pin I10,
The MA mixed fuel pin II11 and the FP extinguishing pin 12 are loaded. Thus, in each segment 8, several types of irradiation fuel pins having different compositions, dimensions, irradiation periods, and the like can be loaded at the same time, and a simulated fuel pin containing no nuclear fuel can be loaded.

As described above, since the irradiation assembly has a more complicated structure than the core fuel assembly, the spatial distribution and the energy distribution of the neutron flux inside the irradiation assembly are more complicated than the core fuel assembly.

In the core calculation for obtaining the effective multiplication factor of the fast reactor core and the output of each fuel assembly, usually, each assembly is completely homogenized in the horizontal direction to obtain a nuclear cross-sectional area, and a homogeneous system is used. Solve the multi-group diffusion equation by the finite difference method. In the finite difference method, the core is divided radially and axially into small volumes called nodes, and the above equations are assumed to be satisfied in each node. By solving this equation, the neutron flux and output of each node can be obtained.

In general, since a fast reactor has a higher neutron energy and a longer mean free path than a thermal neutron reactor such as a light water reactor, it is not necessary to consider the inhomogeneity inside the fuel assembly and a control rod having a strong absorption is required. Except for the inhomogeneity, there is no significant error in core calculation. This also contributes to the fact that the core fuel assembly is made up of many thin fuel pins of a single composition as described above. For this reason,
In the radial cross section of the fuel assembly, the composition of each node is uniform, and the number of node divisions is about 1 to 6 per hexagonal cross section in consideration of calculation time and computer storage capacity. .

[0010] The irradiation assembly has a significantly non-homogeneous structure as compared with the core fuel assembly. However, this irradiation assembly has conventionally ignored its inhomogeneity and has a nuclear cross-sectional area. Have been created. Using the nuclear cross-section created in this way, a three-dimensional diffusion calculation of the reactor core is performed, and the power distribution and combustion characteristics of the irradiation assembly are obtained.

[0011]

As described above, the internal structure of the irradiation assembly is more complicated than that of the core fuel assembly. A fuel having a different fissile material ratio may be loaded in each segment, or a certain segment may not be loaded with any fissile material. The neutron spectrum at each location changes according to the proportion of the fissile material or moderator at that location, so that the neutron spectrum differs from location to location within the irradiation assembly.

However, in the conventional core calculation, the nuclear cross section is obtained on the assumption that the inside of the irradiation assembly is homogeneous. By such treatment, the neutron spectrum inside the irradiation assembly is uniformed, and the reaction rate distribution of each isotope of U and Pu obtained by multiplying the neutron flux by the nuclear cross section for each energy, for example, The distribution of fission and neutron capture rate will differ from the one that reflects the actual heterogeneous structure. In this case, the output of each irradiation fuel pin in the irradiation assembly and the composition change due to combustion cannot be accurately obtained.

[0013] A conventional method for considering the inhomogeneous structure is to perform an inhomogeneous calculation on an aggregate basis in advance, and to calculate the nuclear cross-sectional area of each region in the aggregate based on the obtained neutron flux distribution. It is conceivable to calculate the average nuclear cross section so that the weight and the reaction rate are preserved, and apply this to the core calculation.

Although this method is simple, if the assembly is located in a place where the neutron spectrum change is large in the core, the inhomogeneous calculation for each assembly does not include the entire core in the calculation range. Neutron spectrum is not obtained,
The resulting aggregate-average nuclear cross-section is inaccurate, and even if the aggregate-average nuclear cross-section is accurate, it is only possible to accurately obtain the reaction amount of the aggregate unit, The amount of reaction of each part of the structure is not determined, and detailed nuclear properties of the irradiation assembly cannot be obtained.

On the other hand, as a method for directly obtaining detailed characteristics in the irradiation assembly, a calculation may be considered in which the irradiation assembly is divided into nodes in the core calculation in detail and a non-homogeneous structure is handled. However, the commonly used finite difference method has the same node division even for the core fuel assemblies that occupy most of the core, where the inhomogeneity is small and the node division need not be so detailed.
Calculation time and computer storage capacity are greatly increased.

Further, as the irradiation progresses, the fuel composition changes, and the power distribution changes. Therefore, it is necessary to calculate the combustion of the composition every fixed combustion period, and to perform the three-dimensional core calculation based on the calculation. Repeating the three-dimensional calculation of the entire core of the detailed node for each combustion period as described above requires enormous calculation time and storage capacity of the computer.

The present invention has been made in consideration of the above circumstances, and can calculate a detailed power distribution and a neutron flux distribution in consideration of the inhomogeneity of an irradiation assembly. An object of the present invention is to provide a method and a device for calculating the power distribution of a fast reactor that can be calculated in time.

[0018]

According to a first aspect of the present invention, there is provided a method for calculating a power distribution of a fast reactor, comprising the steps of: using a nuclear cross-sectional area created based on the composition of each fuel assembly in the core of the fast reactor; A three-dimensional neutron diffusion calculation is performed to calculate the neutron flux distribution and the power distribution. Using the axial buckling of each assembly from the three-dimensional neutron diffusion calculation and the fission source distribution of the core fuel assembly, the core horizontal direction is calculated. The two-dimensional neutron diffusion calculation for calculating the neutron flux distribution and the output distribution for each two-dimensional cross section is performed by more detailed node division than in the three-dimensional neutron diffusion calculation.

According to a second aspect of the present invention, there is provided a method for calculating a power distribution of a fast reactor, wherein a neutron flux distribution and a power distribution of the entire core are obtained by using a nuclear cross-sectional area prepared based on a composition of each fuel assembly in the core of the fast reactor. The three-dimensional neutron diffusion calculation is performed, and the boundary condition expressed by the ratio of the neutron flux and the neutron flow on the irradiation assembly surface from the three-dimensional neutron diffusion calculation and the axial buckling of each assembly are calculated. Using, two-dimensional neutron diffusion calculation to calculate the neutron flux distribution and output distribution for each horizontal two-dimensional cross section only for the irradiation assembly,
It is characterized by performing the node division in more detail than in the case of the three-dimensional neutron diffusion calculation.

According to a third aspect of the present invention, there is provided a fast reactor power distribution calculation method, wherein the two-dimensional neutron diffusion calculation includes six or eighteen core fuel assemblies or control rods around the irradiation assembly. The neutron flux distribution and the power distribution are calculated for the partial core region.

According to a fourth aspect of the present invention, there is provided a method for calculating a power distribution of a fast reactor, wherein the neutron flux distribution and the neutron flux distribution for the entire core are determined by using a nuclear cross-sectional area created based on the composition of each fuel assembly in the core of the fast reactor. A three-dimensional neutron diffusion calculation to calculate the power distribution is performed, and the irradiation assembly is calculated using the boundary condition represented by the ratio of the neutron flux to the neutron flux on the irradiation assembly surface from the three-dimensional neutron diffusion calculation. Three-dimensional neutron diffusion calculation to calculate the neutron flux distribution and output distribution only for
It is characterized in that the node division is performed in more detail than in the case of the three-dimensional neutron diffusion calculation for the entire core.

According to a fifth aspect of the present invention, there is provided the fast reactor power distribution calculation method, wherein the three-dimensional neutron diffusion calculation performed subsequent to the three-dimensional neutron diffusion calculation of the entire reactor core includes six irradiation objects around the irradiation assembly. Alternatively, the process is performed on a partial core region including 18 core fuel assemblies or control rods.

According to a sixth aspect of the present invention, there is provided a power distribution calculation method for a fast reactor, wherein a neutron flux distribution and a neutron flux distribution for the entire core are obtained by using a nuclear cross section created based on the composition of each fuel assembly in the core of the fast reactor. Using the three-dimensional neutron diffusion calculation to calculate the output distribution, using the axial buckling of the irradiation assembly from this three-dimensional neutron diffusion calculation and the neutron flux on the surface of the assembly, targeting only the irradiation assembly Perform two-dimensional neutron diffusion calculation to calculate the neutron flux distribution and power distribution for each axial section, the two-dimensional neutron diffusion calculation targeting only the irradiation aggregate, the interior of the irradiation aggregate has a different composition It is divided into a plurality of homogenized regions, and is performed using a combination of analytical solutions in each homogenized region using continuous conditions of neutron flux and neutron flow at each region boundary To.

According to a seventh aspect of the present invention, there is provided a fast reactor power distribution calculating apparatus for calculating a core power distribution history calculating apparatus for calculating a power distribution and its history of the entire core of a fast reactor, and a power distribution and a power distribution of an irradiation assembly. An irradiation aggregate output distribution history calculation device for calculating a history, and control is performed such that the calculation by the core power distribution history calculation device and the calculation by the irradiation aggregate output distribution history calculation device proceed in time synchronization. An overall control device, wherein the core power distribution history calculation device calculates a core cross section of each assembly in the core based on the core composition, and a core power calculation device based on the core cross section. A core power distribution calculator for calculating the distribution, a core cross-section memory for storing the core cross-sectional area of each assembly, a core composition storage for storing the composition of each assembly, and a core power distribution obtained from the core power distribution Neutron flux And a core combustion calculation device that performs a core combustion calculation based on the irradiation aggregate output distribution history calculation device, wherein the irradiation aggregate output distribution history calculation device calculates a nuclear cross-sectional area of each irradiation aggregate based on the irradiation aggregate composition. Irradiation assembly nuclear cross section calculator, irradiation assembly output distribution calculator for calculating irradiation aggregate output distribution based on nuclear cross section, irradiation assembly for storing nuclear cross section of each assembly A nuclear cross-section storage device, an irradiation assembly composition storage device that stores the composition of each assembly, and an irradiation assembly combustion calculation based on the neutron flux distribution obtained from the irradiation assembly output distribution An assembly combustion calculation device, wherein the core power distribution calculation device and the irradiation assembly power distribution calculation device are respectively tertiary of the entire core according to any one of claims 1 to 6. Neutron diffusion calculation and irradiation assembly The overall control device controls the devices so that these calculations are repeated at each predetermined combustion point, thereby calculating the power distribution history of the entire core, The irradiation aggregate is characterized in that a more detailed output distribution history is calculated.

The power distribution calculation apparatus for a fast reactor according to the invention of claim 8 is characterized in that the irradiation assembly composition obtained by the irradiation assembly combustion calculation apparatus and each assembly composition obtained by the core combustion calculation apparatus. Is input to the core cross section calculator to calculate the nuclear cross section, and the calculated core cross section is input to the core power distribution calculator, so that the result of the combustion calculation of the irradiation assembly is output to the entire core. It is characterized in that it is reflected in distribution calculation.

According to a ninth aspect of the present invention, there is provided a power distribution calculating apparatus for a fast reactor, comprising: an irradiation assembly only, or an irradiation assembly and a surrounding six-assembly or 18 which are performed following the three-dimensional neutron diffusion calculation of the entire core. The detailed power distribution calculation for the partial core region including the aggregate was performed using the angular neutron flux calculated from the neutron flow and the neutron flux at the boundary of the aggregate or the partial core region obtained from the three-dimensional neutron diffusion calculation as the boundary condition. It is characterized in that it is performed based on Monte Carlo calculation.

[0027]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS First Embodiment A power distribution calculation method for a fast reactor according to a first embodiment of the present invention will be described below with reference to the drawings.

In the power distribution calculation method according to the first embodiment, first, as shown in FIG. 1, the composition of each fuel assembly in the core of a fast reactor is input (step 1), and based on the input composition, Then, the nuclear cross section of each fuel assembly is calculated (step 2).

Next, a three-dimensional neutron diffusion calculation is performed on the entire core having 6 radial nodes in the assembly (step 3), and the two-dimensional two-dimensional neutron diffusion of each fuel assembly in the core horizontal direction is performed based on the three-dimensional neutron diffusion calculation. The axial buckling for each section and the fission source distribution of the core fuel assembly are input (steps 4 and 5), and two-dimensional neutron diffusion calculation with 54 radial nodes is performed for each two-dimensional section in the horizontal direction of the core. (Step 6).
In the two-dimensional diffusion calculation, the inhomogeneous structure of the irradiation assembly is treated more accurately.

As shown in FIG. 2, the core arrangement of the fast reactor core to be calculated is 81 hexagonal fuel assemblies 1,
It comprises three irradiation assemblies 4, seven control rods 2, and 126 neutron reflectors 3. The distance between the hexagonal faces of the fuel assembly is about 10 cm, and the axial height of the fuel is 60 cm.
It is.

FIGS. 3A, 3B and 3C show cross-sectional structures of a general core fuel assembly 1 and an irradiation assembly 4. As described above, in the irradiation assembly 4 compared with the core fuel assembly 1, several types of irradiation pins having significantly different compositions are loaded, and the non-homogeneity is increased.

FIG. 4A is a radial node division diagram in the three-dimensional diffusion calculation, and reference numeral 13 denotes a radial node in the three-dimensional diffusion calculation of the whole core. FIG.
(B) is an axial node division diagram of the three-dimensional diffusion calculation of the whole core, wherein reference numeral 14 denotes an axial node of the three-dimensional diffusion calculation of the whole core, and reference numerals 15, 16, and 17 denote gas plenum portions, respectively. 3 shows a core fuel portion and an axial reflector portion.

In the three-dimensional diffusion calculation for the entire core, as shown in FIGS. 4A and 4B, the space nodes in each fuel assembly have 6 points in the radial direction, 12 points in the fuel direction in the axial direction, and 12 points in the axial direction. 18 points for the directional neutron reflector and gas plenum, 30 points in the total axial direction, that is, (126 + 81 + 3)
+7) × 6 × 30 = 39060 nodes exist. The composition is uniform in each node, and is homogenized in the radial direction and assigned the same composition inside the irradiation assembly. This is the conventional method, and the power distribution of the core fuel assembly and the irradiation assembly is obtained from the result.

In order to accurately obtain the neutron spectrum and output distribution of the irradiation assembly, it is necessary to increase the number of radial nodes of the assembly to at least 54 in order to incorporate its detailed structure. At the 54 nodes, a model incorporating the structure of the irradiation assembly as shown in FIG. 5 can be used. In FIG. 5, reference numerals 18, 19, 20, 21, and 22 indicate an irradiation core fuel region, an MA mixed fuel region I, and a MA, respectively.
The mixed fuel area II, the FP annihilation area, the structural material and the coolant area are shown.

The interior of the assembly is roughly divided into seven parts corresponding to the arrangement of the irradiation segments, and the same composition is assigned to the detailed nodes belonging to each part. In the part where no fuel is present, the mixed composition of the structural material and the coolant is assigned, and the heterogeneous structure of the aggregate is modeled in more detail than the conventional method. In the neutron diffusion calculation code based on the finite difference method, which is generally used, the size of the nodes is constant at each location.
Doubling, ie, 351540. This results in a significant increase in computer storage capacity and calculation time.

In order to avoid the above problem, in the first embodiment, the axial buckling and the core of each assembly are provided for each core horizontal section based on the three-dimensional diffusion calculation of 6 nodes in the radial direction per assembly. Finding the fission source distribution of the fuel assembly,
Two-dimensional diffusion calculation is performed using them. In this calculation, there are 54 nodes in the radial direction per aggregate as shown in FIG.
The detailed structure of the irradiation assembly is simulated. Although each node has a homogeneous composition, a composition corresponding to the arrangement of the irradiation segments is assigned as described above.

The two-dimensional diffusion calculation is performed on a representative horizontal plane having about 12 points in the axial direction, similarly to the three-dimensional calculation. However, in general, the aggregate structure in the axial direction is relatively uniform, and the neutron flux distribution and output Since the distribution is smoother than in the radial direction, it is easy to obtain a neutron flux distribution and an output distribution in the axial direction with such a score. The calculation node per one axial plane is (126 + 81 + 3 + 7) × 54 = 111718
It is. On 12 planes, the total number of calculation nodes is 140616. This is about one third of the conventional method in which the radial direction is 54 nodes per aggregate in the three-dimensional diffusion calculation of the entire core. In practice, the computation time is even more advantageous because of the faster iterative computation of the fission source as described above. Also, there is an advantage that the computer storage capacity per one calculation is much smaller.

Hereinafter, a calculation formula in the power distribution calculation of the fast reactor described above will be described. The basic equation in the three-dimensional diffusion calculation of the whole core is expressed as follows.

[0039]

(Equation 1)

Here, the symbols are as follows.

[0041]

(Equation 2)

This equation is applied to each node. Each symbol is originally given a subscript of the node, but is omitted here. The same applies hereinafter.

In practice, the first term of this equation is obtained by differentiating the derivative, and the core is divided into nodes (small volumes) to form a finite difference equation and then solved. Since only the basic equations need to be dealt with in the description, the description will proceed in the form of basic equations instead of finite difference equations.

Based on the result of the above calculation, the axial buckling of each fuel assembly and the nuclear fission source distribution of the core fuel assembly in each plane where the two-dimensional calculation is performed are obtained. The axial buckling B _g ² of the g-th group is obtained by the following equation.

[0045]

(Equation 3)

Here, z indicates the coordinate in the axial direction, and Z = Z2
And Z = Z1 indicate the upper and lower surfaces of the axial node of the three-dimensional diffusion calculation to which the horizontal plane to be calculated belongs. Integration regarding Z is performed from Z = Z1 to Z = Z2. The sum regarding j indicates the sum regarding nodes belonging to the same aggregate.

The two-dimensional diffusion calculation subsequent to the three-dimensional diffusion calculation is performed based on the following equation.

[0048]

(Equation 4)

Here, the symbols are as follows. Others are the same as those described above. FS _g ; Fission source of group g obtained by three-dimensional diffusion calculation (fixed value) For the effective multiplication factor, the value of three-dimensional diffusion calculation of the entire core is input. In the two-dimensional diffusion calculation based on this equation,
The node division of the fuel assembly is finer than the three-dimensional diffusion calculation of the entire core, and a detailed non-homogeneous structure of the irradiation assembly is represented.

This makes it possible to calculate more accurate power distribution and neutron flux distribution in the assembly as compared to the conventional method of performing only three-dimensional calculation of the entire core. Further, the calculation time can be greatly reduced as compared with the case where the same detailed node calculation is performed by a three-dimensional system, and the computer storage capacity can be reduced.

In the two-dimensional diffusion calculation, in order to convert a three-dimensional problem into a two-dimensional problem, leakage in the axial direction is handled by a pseudo absorption term using axial buckling. In general, the accuracy of this approximation is good enough because the axial inhomogeneity is small. In addition, since the fission radiation source of the core fuel assembly is a fixed fission radiation source, iterative calculations regarding the fission radiation source are required only for the irradiation assembly, which is advantageous in terms of calculation time.

Although the two-dimensional diffusion calculation is a detailed node, since iterative calculation of the fission radiation source of the core fuel assembly is unnecessary, even if the calculation is repeated for each of the same axial nodes as the three-dimensional diffusion calculation, the same degree is obtained. The calculation time can be greatly reduced as compared with the conventional method of performing detailed node calculation in a three-dimensional system. Since the inhomogeneity of the core fuel assembly is small, the fission source obtained from the three-dimensional diffusion calculation is sufficiently accurate.

Second Embodiment Next, a power distribution calculation method for a fast reactor according to a second embodiment of the present invention will be described with reference to the drawings.

In the output distribution calculation method according to the second embodiment, first, as shown in FIG. 6, the composition of each assembly is input (step 10), and based on the input composition, the composition of each assembly is determined. A nuclear cross section is calculated (step 11).
Next, a three-dimensional neutron diffusion calculation is performed for the entire core having 6 radial nodes per assembly (step 12), and the axial buckling and irradiation surface of each assembly from the three-dimensional neutron diffusion calculation are performed. Input the boundary condition expressed by the ratio of neutron flux to neutron flow (step 13, step 1)
4) The two-dimensional neutron diffusion calculation for each two-dimensional cross section in the horizontal direction targeting only the irradiation assembly
(Step 15). As described above, in the present embodiment, the non-homogeneous structure of the irradiation assembly is more accurately handled.

The irradiation assembly in the two-dimensional diffusion calculation is
The node composition is the same as in FIGS. 3B and 3C. This two-dimensional diffusion equation includes only the irradiation aggregate, and has an advantage that the calculation time is short even if a very detailed node division is used to handle the structure in detail. That is,
Even if the 12 axial planes are repeatedly solved as 54 nodes in the radial direction, the total number of nodes is 54 × 12 = 648, which is about 1/500 the number of nodes compared to the case where this detailed node three-dimensional diffusion calculation is performed for the entire core. There is an advantage that it can be completed.

The calculation formula in the output distribution calculation will be described below. The basic equation in the three-dimensional diffusion calculation of the entire core is expressed as the above-mentioned equation (1). From this calculation result, the axial buckling of the irradiation assembly and the boundary conditions of the surface in each plane where the two-dimensional diffusion calculation for the irradiation assembly is subsequently performed are obtained. The axial buckling is obtained by the equation (3), and the boundary condition is obtained by the following equation for each energy group with respect to each surface serving as an outer boundary in the two-dimensional diffusion calculation.

[0057]

(Equation 5)

Here, the symbols are as follows. Other symbols are as described above. λ _g ; symbol indicating boundary conditions as numerical values (g group) n; normal to the surface that is the outer boundary in two-dimensional diffusion calculation

This boundary condition is a general condition given as an input condition in the three-dimensional diffusion calculation of the whole core. In the three-dimensional diffusion calculation of the whole core, it is applied to the boundary between the core and the outside.
692 are provided. In contrast, this two-dimensional diffusion calculation is applied to the surface of the irradiation assembly,
The difference is that the numerical value of the boundary condition is obtained from the result of the three-dimensional diffusion calculation.

Using these, a two-dimensional diffusion calculation represented by the following equation is performed for each horizontal plane of the irradiation assembly.

[0061]

(Equation 6)

Here, the symbols are as described above.

The calculation range is only the irradiation assembly. In the two-dimensional neutron diffusion calculation, it is possible to simulate the inflow and outflow of neutrons from the outside in a situation where the irradiation aggregate is arranged in the core by the boundary condition. Also,
The leakage in the axial direction is simulated by the pseudo absorption term due to the axial buckling. In this calculation, the non-homogeneous structure of the irradiation assembly is handled using more detailed nodes than in the three-dimensional diffusion calculation, and the fine spatial distribution of the neutron spectrum can be obtained compared to the conventional method, and the detailed output distribution can be calculated with high accuracy. Can be asked well.

Although the two-dimensional diffusion calculation is a detailed node, the calculation time is extremely short because only the irradiation aggregate is included in the calculation model, and the same detailed node calculation is performed in a three-dimensional system. Thus, the calculation time can be greatly reduced, and the computer storage capacity can be reduced.

Third Embodiment Next, a method for calculating the power distribution of a fast reactor according to a third embodiment of the present invention will be described with reference to the drawings.

The third embodiment is different from the above-described second embodiment in that the two-dimensional neutron diffusion calculation is performed by using the irradiation fuel assembly as a center and including the surrounding six or eighteen core fuel assemblies or control rods. It is intended for the core region.

The flow of calculation is the same as in the second embodiment. That is, determine the lambda value lambda _g is the boundary condition of the surface of this portion the core region from the three-dimensional diffusion calculation, inputs the targeting that portion core region, as a boundary condition for two-dimensional diffusion calculation for each axial horizontal plane I do.

FIG. 7A shows a calculation system of two-dimensional diffusion calculation including up to six core fuel assemblies around the irradiation assembly, and FIG. 7B shows a node of the core fuel assembly 1. FIG. 7C shows a node split diagram of the irradiation fuel assembly 4. Reference numerals 18, 19, 2
Reference numerals 0, 21, 22, and 23 denote an irradiation core fuel region, an MA mixed fuel region I, an MA mixed fuel region II, an FP annihilation region, a structural material and a coolant region, and a core fuel region, respectively.

The calculation formula in the present embodiment is the second formula described above.
However, the axial buckling is performed not only for the irradiation aggregate but also for each aggregate to be calculated, and the boundary condition is determined from the three-dimensional diffusion calculation for the outermost aggregate. And apply to two-dimensional diffusion calculations.

As a result, the influence of an error generated when the boundary condition is obtained from the three-dimensional diffusion calculation can be reduced. That is, in the three-dimensional diffusion calculation, the structure of the irradiation assembly is not treated in sufficient detail, so that the boundary condition of the surface may include an error. On the other hand, since the core fuel assembly is sufficient to handle the three-dimensional diffusion calculation, the surface of the partial core region including the six or eighteen core fuel assemblies or control rods around the irradiation assembly is included. It is considered that the boundary conditions are sufficiently accurate.

Therefore, even when an error occurs in the boundary condition of the surface of the irradiation assembly in the three-dimensional diffusion calculation, the fine spatial distribution of the neutron spectrum of the irradiation assembly and the line output of each irradiation fuel can be reduced by this method. It can be obtained with high accuracy. Since the two-dimensional diffusion calculation is limited to the narrow core region including the irradiation assembly, the calculation time is also advantageous.

In the two-dimensional diffusion calculation, the node division of the irradiation assembly is finer than in the three-dimensional diffusion calculation of the whole core, so that the detailed non-homogeneous structure of the irradiation assembly is represented. Has become. Thereby, it is possible to calculate a precise neutron flux distribution and output distribution in an accurate assembly within a short time.

According to the present embodiment, the boundary condition is applied at a position one assembly away from the irradiation assembly. Since the irradiation aggregate is not handled in detail in the three-dimensional diffusion calculation for obtaining the boundary condition, the boundary condition of the surface may have an error. However, in the aggregates other than the irradiation aggregate, the boundary conditions can be obtained with sufficient accuracy even in the three-dimensional diffusion calculation, so that the accuracy of the boundary conditions for the two-dimensional diffusion calculation is improved, and the neutron flux in the irradiation aggregate is improved. The accuracy of distribution / output distribution calculation is improved.

In the present embodiment, as compared with the second embodiment, the calculation range of the two-dimensional diffusion calculation increases from one aggregate to seven aggregates, and the number of nodes in the radial direction increases in proportion to this. Although it increases, it goes without saying that it is advantageous in terms of calculation time as compared with the case where three-dimensional diffusion calculation of the same detail is performed in the entire core.

Fourth Embodiment Next, a method for measuring the power distribution of a fast reactor according to a fourth embodiment of the present invention will be described with reference to the drawings. In the fourth embodiment, in the above-described second or third embodiment, three-dimensional diffusion calculation is performed instead of two-dimensional diffusion calculation only for the irradiation aggregate.

FIG. 8 shows the flow of calculation in the fourth embodiment.
Shown in As shown in FIG. 8, first, the composition of each assembly is input (step 20), and the nuclear cross-sectional area of each assembly is calculated based on the input composition (step 21). next,
A three-dimensional diffusion calculation is performed on the entire core based on the calculated nuclear cross section (step 22).

Next, the boundary conditions on the surface of the irradiation assembly are calculated (step 23). The boundary condition applies the lambda value lambda _g obtained from the core through the 3D diffusion calculation to the surface of the target area.

Finally, a three-dimensional diffusion calculation is performed on the irradiation assembly (step 24). The two-dimensional diffusion calculation is performed for each horizontal section using axial buckling, whereas the three-dimensional diffusion calculation solves the entire axial direction by one calculation. The calculation system of the three-dimensional diffusion calculation including only the irradiation aggregate is the same in the radial direction as that of the second or third embodiment, and the axial direction is the same as that of the three-dimensional diffusion calculation of the whole core, and is 30 nodes Consists of

The advantage of this embodiment is that, as in the second and third embodiments, the number of calculation target aggregates can be limited to one or seven as compared with the case where detailed node division is performed on the entire core. Therefore, the calculation time and the computer storage capacity can be greatly reduced.

The calculation formulas for these calculations will be described below.
The basic equation in the three-dimensional diffusion calculation of the entire core is expressed as the above-mentioned equation (1). From this calculation result, the boundary condition of the irradiation assembly surface is obtained. The boundary condition is obtained by the aforementioned equation (4). However, in this case, it is determined for each three-dimensional surface of the irradiation assembly.

The subsequent three-dimensional diffusion calculation is the same as in the equation (1) except that the above boundary condition is applied to each surface, and only the irradiation aggregate is used. In this three-dimensional diffusion calculation, the node division of the irradiation assembly is finer than in the three-dimensional diffusion calculation of the entire core, and the detailed non-homogeneous structure of the irradiation assembly is expressed. I have.

As a result, it is possible to calculate the detailed neutron flux distribution and power distribution in the irradiation assembly more accurately than in the conventional method, in which the power distribution of the irradiation assembly is obtained from only the three-dimensional diffusion calculation of the entire core. Further, compared to the case where the three-dimensional diffusion calculation of the entire core is performed by the detailed nodes, the storage capacity of the computer can be reduced in a shorter time.

Fifth Embodiment Next, a method for calculating the power distribution of a fast reactor according to a fifth embodiment of the present invention will be described. The fifth embodiment is different from the above-described second or third embodiment in that two-dimensional diffusion calculation is performed instead of the partial core region including the six core fuel assemblies around the irradiation assembly. First, a three-dimensional diffusion calculation is performed. The flow of calculation in the fifth embodiment is the same as the flow of calculation in the fourth embodiment shown in FIG.

The two-dimensional diffusion calculation is performed for each horizontal section using the axial buckling, whereas the three-dimensional diffusion calculation solves the entire axial direction by one calculation. Boundary condition applies the lambda value lambda _g obtained from the core through the 3D diffusion calculation to the surface of the target area. The calculation system of the three-dimensional diffusion calculation including the six core fuel assemblies around the irradiation assembly is the same in the radial direction as in the second and third embodiments, and the axial direction is the entire core. It is the same as in the case of the three-dimensional diffusion calculation and consists of 30 nodes.

The advantage of this embodiment is that, as in the case of the second and third embodiments, the number of calculation target aggregates can be limited to one or seven compared with the case where detailed node division is performed for the entire core. Therefore, the calculation time and the computer storage capacity can be greatly reduced.

Thus, it is possible to suppress the influence of an error generated when the boundary condition is obtained from the three-dimensional diffusion calculation of the entire core. That is, in the three-dimensional diffusion calculation of the entire core, the structure of the irradiation assembly is not dealt with in sufficient detail, so that the boundary condition of the surface may include an error. On the other hand, since the core fuel assemblies need only be handled in the three-dimensional diffusion calculation of the entire core, the boundary of the surface of the partial core region including the six or eighteen core fuel assemblies around the irradiation assembly is sufficient. It is considered that the conditions are sufficiently accurate.

Therefore, even when an error occurs in the boundary condition of the irradiation assembly surface in the three-dimensional diffusion calculation of the entire core, the fine spatial distribution of the neutron spectrum of the irradiation assembly and the line of each irradiation fuel can be obtained by this method. The output can be obtained with high accuracy. Since the three-dimensional diffusion calculation by the detailed nodes is limited to the narrow core region including the irradiation assembly, the calculation time is also advantageous.

Sixth Embodiment A power distribution calculation method for a fast reactor according to a sixth embodiment of the present invention will be described with reference to the drawings.

FIG. 9 shows a calculation flow of the calculation method in this embodiment. As shown in FIG. 9, first, the composition of each assembly is input (step 30), and the nuclear cross-sectional area of each assembly is calculated based on the input composition (step 31).

Next, a three-dimensional neutron diffusion calculation is performed for the entire core having 6 radial nodes per assembly (step 3).
2), input the axial buckling of the irradiation assembly from the three-dimensional neutron diffusion calculation and the neutron flux on the surface of the assembly (step 33), and perform two-dimensional horizontal focusing on only the irradiation assembly. A two-dimensional neutron diffusion calculation is performed for each section (step 34).

The equation for the two-dimensional neutron diffusion calculation in the above heterogeneous system is shown below. When one horizontal cross section of a non-homogeneous irradiation assembly is divided into a plurality of homogenized regions, the following multi-group two-dimensional diffusion equation is established in each of the homogenized regions.

[0092]

(Equation 7)

The symbols are the same as those described above. However,
Each quantity here is defined not as a node having a finite size but as a continuous quantity on a two-dimensional plane representing a cross section of the aggregate.

By solving this equation, the neutron flux of the g-th group is expressed as follows.

[0095]

(Equation 8)

The integration regarding S is performed on the region boundary, and the integration regarding r is performed on the inside of the region. The symbols are as follows.

[0097]

(Equation 9)

Here, G _g (r, S) represents a g-th group neutron flux generated at the point r by the unit source at the point S.

The source σ _g (s) on the outer boundary is determined so as to satisfy the following equation.

Here, at the outer boundary of the assembly, the boundary source σ can be determined by the following equation (8) using the neutron flux on the outer boundary of the assembly obtained from the three-dimensional diffusion calculation of the entire core.

[0101]

(Equation 10)

On the other hand, the internal boundary source σ of the aggregate can be determined from the respective continuity conditions in consideration of the neutron flow in addition to the neutron flux at the boundary. The formula for calculating the neutron flow is shown below.

[0103]

[Equation 11]

Here, each amount is as follows.

[0105]

(Equation 12)

The neutron flow on the boundary is calculated by the following equation.

[0107]

(Equation 13)

Finally, the continuity formulas for the neutron flux and the neutron flow are shown below.

[0109]

[Equation 14] Here, s0 + and s0- indicate a case where the limit is taken from the right side and a case where the limit is taken from the left with respect to the point s0 on the internal boundary, respectively. Specifically, it shows a neutron flow defined from the right region of the internal boundary, and a neutron flow and neutron flux defined from the left region.

The integration in the above equations (6) to (12) is actually performed by numerically integrating the area boundary and the interior into several small parts.

Since the internal boundary source has different values on the left and right of the boundary, it has two values for one point on the boundary.
Since as many equations as unknowns are required to determine these internal boundary sources, equations for the neutron flux and neutron flow continuity are imposed. As a result, the analytical solutions in the respective homogenized regions can be combined, and the neutron flux distribution and output distribution can be determined by inputting the neutron flux at the outer boundary of the assembly.

According to the above calculation method, the neutron flux distribution and the output distribution at an arbitrary position can be obtained with high accuracy while treating the irradiation aggregate in a non-uniform manner. This is because the calculation of the neutron flux distribution and output distribution by this method is not performed for each node as in the finite difference method, but based on an analytical solution, and the neutron flux distribution and output distribution are continuous in space. It is because it is obtained. Further, in the present method, the calculation time is short because it is based on the analytical solution, and the calculation time can be reduced as compared with the case where the entire core is treated inhomogeneously with detailed nodes.

FIG. 3 (B) loaded in the fast reactor core of FIG.
FIG. 10 shows a calculation system of the two-dimensional diffusion calculation for the irradiation assembly shown in FIG. Reference numerals 9, 10, 11, and 12 in the figure denote irradiation core fuel region, MA mixed fuel pin region I, respectively.
The MA mixed fuel pin region II and the FP disappearance region are shown.
The interior of this system consists of seven homogenized regions corresponding to the irradiation segments.

In each region, a neutron flux distribution calculation formula based on an analytical solution in which the boundary source is unknown is established. To obtain an external source, a neutron flux from a three-dimensional core calculation is used. To determine the source, the neutron flux and neutron flux continuity conditions at the boundary are imposed, and the neutron flux distribution and output distribution of the entire assembly can be determined.

In this calculation, the non-homogeneous structure of the irradiation assembly is treated more accurately than the three-dimensional diffusion calculation of the entire reactor core, and it is based on an analytical solution. Neutron flux distribution and power distribution can be obtained.

As described above, according to the present embodiment, the neutron flux and power distribution of the entire core are calculated, and detailed calculations are performed for the irradiation aggregate having a large degree of inhomogeneity. The neutron flux distribution and output distribution of the aggregate can be obtained.

Seventh Embodiment Next, a fast reactor power distribution calculating apparatus according to a seventh embodiment of the present invention will be described with reference to the drawings. FIG. 11 shows the configuration of the computing device in this embodiment. In this apparatus, a combustion calculation for each detailed node is performed based on the neutron flux distribution from the irradiation aggregate output distribution calculation according to the above-described second embodiment, and this is repeated for each predetermined combustion period.

The configuration and function of this apparatus will be described with reference to an example of the case of the fast reactor core shown in FIG. In this fast reactor, one cycle between fuel exchanges is about 90 days, and the core fuel assembly and the irradiation assembly shown in FIGS. 3A and 3B are taken out after six cycles of loading. . That is, it is loaded for 90 × 6 days. In order to track changes in the composition and power of the irradiation assembly during this time, power distribution calculation and combustion calculation for each detailed node of the irradiation assembly are performed along with power distribution calculation and combustion calculation for the entire core.

The fast reactor power distribution calculation device in FIG. 11 includes a core power distribution history calculation device 24, an irradiation assembly power distribution history calculation device 25, a core cross section storage device 26, a core composition storage device 27, an irradiation assembly Body cross section storage device 2
8, irradiation assembly composition storage device 29, overall control device 30
Consists of

The core power distribution history calculating device 24 calculates the history of the power distribution of the entire core, and comprises a core core area calculating device 31, a core power distribution calculating device 32, and a core burning calculating device 33. The irradiation aggregate output distribution history calculating device 25 calculates the output distribution of the irradiation aggregate and its history. The irradiation aggregate core cross-sectional area calculation device 34, the irradiation aggregate output distribution calculation device 35, It consists of an assembly combustion calculator 36 for use. The overall control device 30 performs control so that the core power distribution history calculation and the irradiation aggregate output distribution history calculation progress in time synchronization.

The core core area calculation unit 31 reads the core composition (initial state) from the core composition storage unit 27, calculates the core cross section of each assembly in the core, and stores the calculation result in the core core area storage unit 26. To be stored. Next, the core power distribution calculator 32 calculates the core power distribution using the core core cross-sectional area.
The obtained neutron flux distribution of the core together with the core composition from the core composition storage device 27 and the core core cross-sectional area from the core core cross-section storage device 26 is input to the core combustion calculator 33, where the combustion calculation is performed. Will be As a result, the core composition after burning for a certain time is obtained and stored in the core composition storage device 27. This new core composition is input to the core cross-sectional area calculation device 31, and the power distribution after burning for a certain period of time is calculated by the same flow as above.

By repeating this, the core power distribution history is calculated. The irradiation aggregate output distribution history calculation device 25 has the same configuration, and the irradiation aggregate output distribution history is calculated in a similar flow.

The core power distribution calculation device 32 calculates the power distribution by three-dimensional diffusion calculation.
As shown in (B), the number of nodes for calculating the neutron flux distribution and the output distribution is six (radial direction) per aggregate. The combustion calculation is performed with the neutron flux obtained by averaging the six nodes shown in FIG. Accordingly, the core composition and the nuclear cross-sectional area calculated based on the core composition are treated as an aggregate unit, that is, an aggregate average in the radial direction. For this reason, the core cross-sectional area storage unit 26 and the core composition storage unit 27 store the average core cross-sectional area and composition in the radial direction.

From the core power distribution calculation device 32,
The axial buckling for each axial section of the irradiation assembly divided into 5 cm sections and boundary conditions given as the ratio of neutron flow / neutron flux on the surface of the assembly are output, and the irradiation assembly output distribution calculator 35 is output. Is input to In the core power distribution calculation device 32, the power distribution is naturally calculated for the irradiation assembly as one assembly in the core, but the node division is treated in the same manner as the radial direction 6 and the core fuel.

The irradiation aggregate output distribution calculating device 35 calculates the output distribution in the irradiation aggregate divided into 54 nodes as shown in FIG. 5 for each axial section sectioned at every 5 cm of the irradiation aggregate. Calculated by two-dimensional diffusion calculation. This is more detailed than the 6-node assembly in the core power distribution calculation device 32, and the power distribution can be calculated in detail by taking into account the heterogeneity of the irradiation assembly.

Further, the neutron flux of the detailed node is input to the irradiation aggregate combustion calculating device 36, and the irradiation aggregate composition and the irradiation aggregate core cross-sectional area based on the composition are calculated for each detailed node. Therefore, in the irradiation assembly composition storage device 29 and the irradiation assembly core cross section storage device 28,
In each case, the irradiation aggregate composition and the nuclear cross-sectional area are 5 in the radial direction.
It is stored for each of the four detailed nodes. This is a difference from the core composition storage device 27 and the core cross-sectional area storage device 26 that store only the average value of the aggregates. Thus, both the power distribution and the combustion composition of the core power distribution history calculation device 2
4 Calculated in more detail.

The calculation of the core power distribution history calculation device 24 and the irradiation assembly power distribution history calculation device 25 is performed by the overall control device 30 in synchronization with time. FIG. 12 shows the flow of this calculation. By repeating the 90-day combustion for 6 cycles, the history of the detailed output distribution until the irradiation aggregate is taken out can be tracked.

As shown in FIG. 12, in the power distribution calculation apparatus according to the present embodiment, the core area calculation (step 40), the core power distribution calculation (step 41), the core combustion calculation (step 42), the core The core power distribution is calculated in the order of the cross-sectional area calculation (step 43) and the core power distribution calculation (step 44), the core cross-sectional area calculation for the irradiation assembly (step 45), the detailed power distribution calculation for the irradiation assembly (step 45). Step 46), calculation of the fuel pin combustion of the irradiation assembly (step 47), calculation of the irradiation assembly nuclear cross section (step 48), and calculation of the detailed output distribution for the irradiation assembly (step 49). Compute the aggregate output distribution. In the above calculation, the result of the core power distribution calculation (steps 41 and 44) is reflected in the irradiation aggregate power distribution calculation (steps 46 and 49).

Since the neutron flux distribution changes greatly within the irradiation assembly, the output distribution based on the three-dimensional diffusion calculation of the whole core, which is not a detailed node, has a large error, as in the conventional method, and the neutron flux distribution is used. In the combustion calculation, the error of the combustion composition becomes large. Further, performing the detailed node three-dimensional diffusion calculation for all the cores and performing the combustion calculation based on the calculation requires enormous calculation time and computer storage capacity.

On the other hand, in the calculation device according to the present embodiment, the output distribution of the irradiation assembly is calculated by the output distribution calculation method according to any of the second to sixth embodiments.
Combustion calculation by detailed neutron flux distribution shows that
Or, it is limited to 6 or 18 bodies in the vicinity. By doing so, it is possible to greatly reduce the time required and to reduce the storage capacity of the combustion composition and the nuclear cross section as compared with the case where the combustion calculation of the entire core is performed at the detailed node.

As a result, the fine spatial distribution of the neutron spectrum in the irradiation assembly and the linear output of each irradiation fuel at each combustion point can be accurately obtained in a short time. Further, there is an advantage that the calculation of the combustion based on the detailed neutron flux distribution can be performed separately from the calculation of the three-dimensional diffusion of the entire core, and can be performed as needed.

According to the output distribution calculating apparatus of the present embodiment, the aggregate for obtaining the detailed output distribution inside the aggregate and performing the combustion calculation for each detailed node can be limited to only the irradiation aggregate. As compared with the case where a detailed power distribution is obtained over the entire core and the combustion calculation is repeated as in the case of the conventional apparatus, the present embodiment is significantly advantageous in terms of the capacity of the storage device and the calculation time.

Eighth Embodiment Next, a description will be given of an apparatus for calculating a power distribution of a fast reactor according to an eighth embodiment of the present invention with reference to the drawings. FIG. 13 shows the configuration of the computing device in this embodiment. In the calculation device according to the present embodiment, similarly to the calculation device according to the seventh embodiment, the core power distribution history calculation and the irradiation aggregate output distribution history calculation are performed. In the latter calculation, the calculation node is more detailed than the former calculation. , A detailed output distribution history of the irradiation assembly can be accurately obtained in a short time.

This embodiment is different from the seventh embodiment in that the composition of the irradiation assembly after combustion is input to the core power distribution history calculation device 24 as a part of the core composition, and the detailed node Is to reflect the composition of the irradiation assembly to the core power distribution history calculation.

More specifically, the irradiation assembly combustion calculator 36 is used.
Is stored in the core composition storage device 27, and is reflected in the calculation of the core cross section, the calculation of the core power distribution, and the calculation of the core combustion. Here, the composition of each detailed node is averaged so as to match the node of the power distribution calculation of the entire core. FIG. 14 shows the flow of this calculation.

As shown in FIG. 14, in the power distribution calculating apparatus according to the present embodiment, the core area calculation (step 50), the core power distribution calculation (step 51), the core combustion calculation (step 52), the core core The core power distribution is calculated in the order of the cross-sectional area calculation (step 53) and the core power distribution calculation (step 54), the core cross-sectional area calculation for the irradiation assembly (step 55), and the detailed power distribution calculation for the irradiation assembly (step 55). Step 56), calculation of the fuel pin combustion of the irradiation assembly (step 57), calculation of the irradiation assembly core cross-sectional area (step 58), and calculation of the detailed output distribution for the irradiation assembly (step 59). Compute the aggregate output distribution.

In the above calculation, the result of the core power distribution calculation (steps 51 and 54) is reflected on the irradiation assembly power distribution calculation (steps 56 and 59), and each fuel pin of the irradiation assembly is Combustion calculation (Step 5
The result of 7) is reflected in the calculation of the core cross section (step 53).

The functions and effects obtained by the power distribution calculating apparatus according to the present embodiment are almost the same as those of the power distribution calculating apparatus according to the seventh embodiment. The difference is that it is reflected in the three-dimensional diffusion calculation of the entire core, and this makes it possible to incorporate the influence of the detailed location-specific combustion of the irradiation assembly on the power distribution of the entire core. The neutron flux and power distribution inside the irradiation assembly vary greatly in combustion, and the neutron flux and
When the power distribution is affected, the combustion history of the power distribution can be tracked with higher accuracy than the power distribution calculation device according to the seventh embodiment.

Also in this embodiment, the output distribution can be obtained for each detailed node, and the aggregate for performing the combustion calculation can be limited to only the irradiation aggregate. Compared with the case where the combustion calculation is repeated, it is significantly advantageous in terms of computer capacity and calculation time. When the neutron flux and power distribution inside the irradiation assembly greatly change in combustion and affect the neutron flux and power distribution of the entire core, the reactor core is more accurately compared to the power distribution calculator of the seventh embodiment. Output distribution history can be tracked.

Ninth Embodiment Next, a description will be given of a fast reactor power distribution calculating apparatus according to a ninth embodiment of the present invention with reference to the drawings. The configuration of the computing device in this embodiment is the same as the configuration shown in FIG. However, the difference is that the irradiation aggregate output distribution calculation device 35 is not a diffusion calculation of detailed nodes but a Monte Carlo calculation. FIG. 15 shows the flow of this calculation.

As shown in FIG. 15, in the power distribution calculating apparatus according to the present embodiment, the core cross-sectional area calculation (step 60), the power distribution calculation of the whole core (step 61),
Core burnup calculation (step 62), core core cross section calculation (step 63), power distribution calculation of the entire core (step 6)
The core power distribution is calculated in the order of 4), the core cross section of the irradiation assembly is calculated (step 65), the Monte Carlo detailed power distribution calculation for the irradiation assembly (step 66),
Detailed node combustion calculation of irradiation assembly (step 67),
The irradiation assembly output distribution is calculated in the order of the irradiation assembly core cross-sectional area calculation (step 68) and the Monte Carlo detailed output distribution calculation (step 69) for the irradiation assembly.

In the above calculation, the result of the power distribution calculation of the entire core (steps 61 and 64) is used as the boundary condition to calculate the irradiation assembly output distribution (steps 66 and 6).
This is reflected in 9).

In the power distribution calculation apparatus according to this embodiment, three-dimensional diffusion calculation of the entire core and combustion calculation of each assembly composition are performed every 60 days in the fast reactor core of FIG. For the irradiation assembly, Monte Carlo calculations are performed using the detailed assembly structure with the neutron flux calculated from the three-dimensional diffusion calculation of the entire core and the neutron flux calculated from the neutron flow as the boundary conditions. In addition, the calculation of combustion is performed every 60 burning days using the obtained neutron flux.

According to the calculation method and the calculation apparatus, the history of the three-dimensional power distribution of the entire core is calculated in the conventional manner.
The Monte Carlo calculation can be applied only to the irradiation aggregates whose aggregate structure is complicated and need to be handled in detail.
It is possible to perform highly accurate detailed output distribution calculation in a short time.

In the Monte Carlo calculation, the behavior of neutrons can be treated with almost no approximation based on the transport equation, and the system can be treated non-homogeneously with almost no approximation. Calculation of distribution and output distribution is possible. However, obtaining a detailed power distribution for the irradiation assembly while solving the entire core requires an extremely enormous amount of calculation time.

Therefore, the calculation of the entire core is carried out by the conventional three-dimensional diffusion calculation, and is limited to the irradiation aggregate for which detailed characteristics are to be obtained, or to the irradiation aggregate and the aggregate in the vicinity thereof. By performing the Monte Carlo calculation
Detailed calculation of neutron flux distribution and power distribution of the irradiation assembly can be performed accurately and efficiently. For that purpose, it is effective to pass the boundary condition represented by the following formula to the Monte Carlo calculation from the three-dimensional diffusion calculation of the entire core.

[0147]

(Equation 15)

Here, each amount is as follows.

[0149]

(Equation 16)

The above-mentioned angular neutron flux is defined on each surface of the aggregate which is a boundary between the three-dimensional diffusion calculation and the Monte Carlo calculation, and is calculated based on the result of the three-dimensional diffusion calculation. The angular neutron flux is used because the Monte Carlo calculation basically deals with the angular neutron flux for each neutron flight direction.

With this calculation device, the three-dimensional calculation of the whole core is performed based on the conventional diffusion calculation, and Monte Carlo calculation with high approximation accuracy can be applied only to an assembly requiring a detailed model. It is possible to obtain a precise neutron flux distribution and output distribution in a very short time in comparison with the case where Monte Carlo calculation is applied to the neutron flux.
In this case, from the three-dimensional calculation of the entire core, to pass the boundary conditions as an angular neutron flux indicating the inflow and outflow of neutrons to the aggregate for which you want to obtain a detailed neutron flux distribution and power distribution,
It is possible to accurately simulate a state where the target assembly is loaded on all the cores.

Further, by performing the combustion calculation using the neutron flux of the irradiation fuel pin obtained based on this calculation device, the combustion history of the output distribution of the irradiation assembly can be accurately obtained.

In the first to ninth embodiments described above, the irradiation aggregate containing a fissile material is described. However, for example, a core material such as a moderator / shielding material as shown in FIG. 16 is used. Similarly, the irradiation assembly including the control rod assembly loaded with the absorber as shown in FIG. 17 also has a large inhomogeneity in the internal structure, and the calculation method according to each of the above embodiments is also used for such an assembly. By applying the computer and the calculation device, the detailed neutron flux distribution and power distribution and the combustion history thereof can be accurately obtained in a short time.

In FIG. 16, reference numerals 37, 38, 3
Reference numeral 9 denotes a hydrogenated zirconium pin, an improved stainless steel pin, and a beryllia pin. In FIG. 17, reference numerals 40, 41, and 42 denote an absorber pin, a guide tube,
The protection tube is shown.

The power distribution calculation method and calculation device of the fast reactor according to the first to ninth embodiments of the present invention have been described above. The effects of these embodiments will be listed below. (1) By using a model that deals with the structure inside the irradiation assembly of the fast reactor in more detail than the conventional method, a fine spatial distribution of the neutron spectrum in the irradiation assembly can be obtained, and the The line output can be obtained with high accuracy. (2) In addition, the calculation time can be significantly reduced as compared with the case of the conventional method which treats an equivalent heterogeneous structure.
A large increase in computer storage capacity can be suppressed. (3) By performing combustion calculation using the obtained neutron spectrum for each irradiation detail node, it is possible to accurately predict the composition change and output history due to combustion of each irradiation fuel pin in a short time, and to perform irradiation. It is possible to accurately evaluate the life of the fuel pin for use. (4) Since the thermal margin and life of the irradiation assembly can be accurately predicted, efficient and accurate irradiation can be performed.

[0156]

As described above, according to the method and apparatus for calculating the power distribution of a fast reactor according to the present invention, a fine spatial distribution of the neutron spectrum in the irradiation assembly can be obtained, and the The line output can be obtained with high accuracy. In addition, the calculation time can be significantly reduced as compared with the case of the conventional method that treats an equivalent heterogeneous structure, and a large increase in computer storage capacity can be suppressed.

[Brief description of the drawings]

FIG. 1 is a calculation flowchart of a power distribution calculation method for a fast reactor according to a first embodiment.

FIG. 2 is a configuration diagram showing an example of a fast reactor core.

3A is a configuration diagram of a core fuel assembly, FIG. 3B is a configuration diagram of an irradiation assembly, and FIG. 3C is a table showing specifications of each fuel pin.

FIGS. 4A and 4B are radial and axial node division diagrams in three-dimensional diffusion calculation, respectively.

FIG. 5 is a detailed node division diagram of the irradiation aggregate in the two-dimensional diffusion calculation.

FIG. 6 is a calculation flowchart of a power distribution calculation method for a fast reactor according to the second embodiment.

7A is a diagram showing a calculation system of two-dimensional diffusion calculation, and FIGS. 7B and 7C are node division diagrams of a core fuel assembly and an irradiation fuel assembly, respectively.

FIG. 8 is a calculation flowchart of a power distribution calculation method for a fast reactor according to a fourth embodiment.

FIG. 9 is a calculation flowchart of the power distribution calculation device for a fast reactor according to the sixth embodiment.

FIG. 10 is a calculation system diagram of an irradiation assembly according to the sixth embodiment.

FIG. 11 is a configuration diagram of a fast reactor power distribution calculation device according to a seventh embodiment.

FIG. 12 is a calculation flowchart of a fast reactor power distribution calculation device according to a seventh embodiment.

FIG. 13 is a configuration diagram of a fast reactor power distribution calculation device according to an eighth embodiment.

FIG. 14 is a calculation flowchart of the power distribution calculation device for the fast reactor according to the eighth embodiment.

FIG. 15 is a calculation flowchart of the power distribution calculation device for the fast reactor according to the ninth embodiment.

FIG. 16 is a configuration diagram showing an example of a core material irradiation assembly.

FIG. 17 is a configuration diagram showing an example of a control rod assembly.

[Explanation of symbols]

 REFERENCE SIGNS LIST 1 core fuel assembly 2 control rod 3 neutron reflector 4 irradiation assembly 6 core fuel pin 7 wrapper tube 8 irradiation segment 9 irradiation core fuel pin 10 minor actinide mixed fuel pin I 11 minor actinide mixed fuel pin II 12 FP Extinguishing pin 13 Radial node for three-dimensional diffusion calculation of entire core 14 Axial node for three-dimensional diffusion calculation of entire core 15 Gas plenum 16 Core fuel 17 Axial reflector 18 Irradiation core fuel area 19 Minor actinide mixed fuel pin area I 20 Minor actinide mixed fuel pin region II 21 FP annihilation region 22 Structural material and coolant region 23 Core fuel region 24 Core power distribution history calculation device 25 Irradiation assembly power distribution history calculation device 26 Core core cross-sectional area storage device 27 Core Composition storage device 28 Irradiation assembly nuclear cross-section area storage device 2 Reference Signs List 9 Irradiation assembly composition storage device 30 Overall control device 31 Core core cross-section calculation device 32 Core power distribution calculation device 33 Core combustion calculation device 34 Irradiation assembly core cross-section calculation device 35 Irradiation assembly output distribution calculation device 36 Irradiation assembly combustion calculator

Claims (9)

[Claims] 1. A three-dimensional neutron diffusion calculation for calculating a neutron flux distribution and a power distribution over the entire core using a nuclear cross section created based on a composition of each fuel assembly in a core of a fast reactor, Two-dimensional neutron diffusion calculation to calculate neutron flux distribution and power distribution for each two-dimensional cross section in the core horizontal direction using axial buckling of each assembly from neutron diffusion calculation and fission source distribution of core fuel assembly Is performed by more detailed node division than in the case of the three-dimensional neutron diffusion calculation. 2. A three-dimensional neutron diffusion calculation for calculating a neutron flux distribution and a power distribution of the entire core using a nuclear cross section created based on the composition of each fuel assembly in the core of the fast reactor, Using the boundary condition expressed by the ratio of neutron flux and neutron flow on the surface of the irradiation assembly from the neutron diffusion calculation and the axial buckling of each assembly, the horizontal A power distribution calculation method for a fast reactor, wherein a two-dimensional neutron diffusion calculation for calculating a neutron flux distribution and a power distribution for each two-dimensional cross section is performed by more detailed node division than in the three-dimensional neutron diffusion calculation. 3. The two-dimensional neutron diffusion calculation calculates a neutron flux distribution and a power distribution in a partial core region including six or eighteen core fuel assemblies or control rods around the irradiation assembly. 3. The method according to claim 2, wherein the power distribution is calculated. 4. A three-dimensional neutron diffusion calculation for calculating a neutron flux distribution and a power distribution for the entire core by using a nuclear cross section created based on the composition of each fuel assembly in the core of the fast reactor, Using a boundary condition expressed by the ratio of neutron flux to neutron flow on the surface of the irradiation assembly from three-dimensional neutron diffusion calculation, the cubic calculation of the neutron flux distribution and power distribution for only the irradiation assembly A power distribution calculation method for a fast reactor, wherein the source neutron diffusion calculation is performed by more detailed node division than in the case of the three-dimensional neutron diffusion calculation for the entire core. 5. The three-dimensional neutron diffusion calculation performed subsequent to the three-dimensional neutron diffusion calculation of the whole core includes a part including the core fuel assemblies or control rods of six or eighteen surroundings around the irradiation assembly. 5. The method according to claim 4, wherein the method is performed for a core region. 6. A three-dimensional neutron diffusion calculation for calculating a neutron flux distribution and a power distribution for the entire core using a nuclear cross section created based on the composition of each fuel assembly in the core of the fast reactor, Using the axial buckling of the irradiation assembly from the three-dimensional neutron diffusion calculation and the neutron flux on the surface of the assembly, calculate the neutron flux distribution and output distribution for each axial section only for the irradiation assembly Perform two-dimensional neutron diffusion calculation to, two-dimensional neutron diffusion calculation only for the irradiation assembly, the interior of the irradiation assembly is divided into a plurality of homogenized regions having different compositions, at each region boundary A method for calculating the power distribution of a fast reactor, wherein the method is performed using a combination of analytical solutions in respective homogenization regions using continuous conditions of neutron flux and neutron flow. 7. A core power distribution history calculating device for calculating the power distribution and its history of the entire core of a fast reactor, an irradiation assembly power distribution history calculating device for calculating the irradiation distribution and its history. An overall control device that controls the calculation by the core power distribution history calculation device and the calculation by the irradiation aggregate output distribution history calculation device to proceed in synchronization with time, the core power distribution history calculation The apparatus includes a core nuclear cross section calculator that calculates a nuclear cross section of each assembly in the core based on the core composition, a core power distribution calculator that calculates a core power distribution based on the nuclear cross section, and each assembly. Core cross-section storage device that stores the nuclear cross-sectional area of the core, core composition storage device that stores the composition of each assembly, and core combustion calculation that performs the core combustion calculation based on the neutron flux distribution obtained from the core power distribution Equipment and The irradiation assembly output distribution history calculation device, the irradiation assembly core cross section calculation device for calculating the nuclear cross section of each irradiation assembly based on the irradiation assembly composition, the nuclear cross section An irradiation assembly output distribution calculation device that calculates an irradiation aggregate output distribution based on the irradiation assembly nuclear cross section storage device that stores a nuclear cross section of each assembly; and stores a composition of each assembly. An irradiation assembly composition storage device, and an irradiation assembly combustion calculation device that performs irradiation assembly combustion calculation based on a neutron flux distribution obtained from the irradiation assembly output distribution, wherein the core power distribution The calculation device and the irradiation assembly output distribution calculation device each include a three-dimensional neutron diffusion calculation for the entire core and a detailed output distribution calculation for the irradiation assembly according to any one of claims 1 to 6. The overall control device performs this The above devices are controlled such that the calculation of the above is repeated for each predetermined combustion time point, whereby the power distribution history of the entire core is calculated, and the power distribution history of the irradiation assembly is calculated in more detail. A power distribution calculation device for a fast reactor, characterized in that: 8. An irradiation assembly composition obtained by the irradiation assembly combustion calculator and each assembly composition obtained by the core combustion calculator are inputted to a core core cross section calculator and a nuclear cross section is obtained. The calculated nuclear cross section is input to the core power distribution calculation device, so that the result of the combustion calculation of the irradiation assembly is reflected in the power distribution calculation of the entire core. Item 8. A power distribution calculation device for a fast reactor according to Item 7. 9. The detailed power distribution calculation for the irradiation core only or the irradiation core and a partial core region including the surrounding six or eighteen assemblies following the three-dimensional neutron diffusion calculation of the whole core is performed. The method is performed based on Monte Carlo calculation using a neutron flux calculated from a neutron flow and a neutron flux at a boundary of the aggregate or a partial core region obtained from a three-dimensional neutron diffusion calculation as a boundary condition. The power distribution calculation device for a fast reactor according to claim 7 or 8.