JP5199557B2 - 3D core analysis method and 3D core analysis program - Google Patents

Description

  The present invention relates to a three-dimensional core analysis method and a three-dimensional core analysis program for analyzing, evaluating and managing the power distribution and reactivity of a reactor core having a plurality of fuel rods by numerical calculation.

  Conventionally, there has been known a method of performing core management by obtaining effective multiplication factor and neutron flux distribution of a neutron in a nuclear reactor core by numerical analysis, evaluating characteristic data such as an effective multiplication factor and power distribution (Patent Document 1). , 2 etc.).

  Examples of such reactor core analysis methods include a diffusion calculation method based on a neutron diffusion theory, a transport calculation method based on a neutron transport theory, and a Monte Carlo method which is a stochastic method.

On the other hand, in the development of nuclear fuel, the development of highly heterogeneous cores, such as the use of high burnup fuel with a maximum burnup exceeding 55 GWd / t, is expected in the future. Development of a core analysis method is desired.
JP 2005-227174 A Japanese Patent No. 3245640

  However, in order to perform highly accurate core analysis for a three-dimensional inhomogeneous geometric system, it is desirable to perform neutron transport calculations that take into account the multidirectional nature of neutron motion. Becomes enormous and analysis must be performed over a long period of time on a computer with a large storage capacity.

  Therefore, an object of the present invention is to provide a three-dimensional core analysis method and a three-dimensional core analysis program that can perform accurate core analysis while reducing the load on the computer.

  In order to achieve the above-mentioned object, the three-dimensional core analysis method of the present invention provides the behavior of neutrons in a three-dimensional cell in a system in which a core having a plurality of fuel rods is divided into a plurality of three-dimensional solid cells. A three-dimensional core analysis method for analyzing a two-dimensional surface of the three-dimensional cell, wherein a neutron track passing through the surface at a certain azimuth angle ψ is defined by a plurality of linear paths, and the linear path is defined as the two-dimensional surface. A surface region extending in a direction perpendicular to the surface is generated, and the surface neutron flux of the surface region is calculated from the neutron flux incident at the elevation angle θ and the emitted neutron flux in the surface of the surface region. By calculating the neutron flux for all the surface areas on the plurality of straight paths, the angular neutron flux of the solid cell is calculated, and the same calculation is performed for the plurality of azimuth angles Ψ and elevation angles θ. All angles And calculates the three-dimensional neutron flux of the solid cell by angle integrated to calculate a child bundle.

  Here, the surface area can be a rectangular area having a line segment divided by a point where the straight line path intersects a dividing line arbitrarily dividing the two-dimensional surface of the three-dimensional cell as an upper side or a bottom side. .

  Further, the distribution of the incident neutron flux can be input based on a functional expression.

  Furthermore, the three-dimensional core analysis program of the present invention uses a computer to analyze neutron behavior in a three-dimensional cell in a system in which a core having a plurality of fuel rods is divided into a plurality of three-dimensional solid cells. A three-dimensional core analysis program to be operated, wherein a plurality of parallel means for inputting tracks for analyzing conditions and paralleling tracks of neutrons passing through a two-dimensional plane of the solid cell at a certain azimuth angle Ψ based on the analysis conditions The azimuth angle ψ is changed to generate a straight path for each azimuth angle ψ, and the straight path is extended in a direction perpendicular to the two-dimensional plane to obtain a plane region p. The surface neutron flux φ (p, θ, Ψ) in the surface region from the surface region generating means for generating the neutron flux incident at the elevation angle θ and the emitted neutron flux in the plane of the surface region p. Calculate for the straight path In addition, by integrating the plane neutron flux φ (p, θ, Ψ) of the plane region p included in the space of the three-dimensional cell in the space, the angular neutron of the three-dimensional cell toward the elevation angle θ and the azimuth angle Ψ Angular neutron calculation means for calculating the bundle φ (θ, Ψ) and neutron flux calculation for calculating the three-dimensional neutron flux φ in the three-dimensional cell by integrating the angular neutron flux φ (θ, Ψ) with the angle. And means for outputting an analysis result based on the three-dimensional neutron flux φ.

  In the three-dimensional core analysis method of the present invention thus configured, a three-dimensional solid cell is two-dimensionalized, a plurality of straight paths are defined for the two-dimensional plane, and the straight paths are orthogonal to the two-dimensional plane. The components in the three directions are secured by generating a surface region orthogonal to the two-dimensional surface by extending in the direction.

  Then, a plane neutron flux is calculated from a neutron flux incident on a surface area developed from a two-dimensional plane linear path and an emitted neutron flux, and a three-dimensional neutron flux is calculated by integrating them.

  By adopting a method of developing a straight path defined on a two-dimensional surface in this way, it is possible to suppress an increase in the amount of calculation and data, and to perform highly accurate three-dimensional core analysis in a short time. .

  In other words, if all the tracks of neutrons that pass through a three-dimensional cell are individually set in a three-dimensional space and analysis is performed based on the tracks, the amount of calculations, input parameters, mesh data, etc. will be enormous. Therefore, even if a computer having a high processing speed and a large storage capacity is used, the analysis must be performed for a long time.

  On the other hand, by adopting the configuration of the present invention as described above, the load on the computer is small and accurate core analysis can be performed even in three dimensions.

  In addition, increasing the number of divisions of the straight path increases the calculation time, but the influence from the boundary surface of the analysis region can be reduced and the accuracy can be improved, so that the straight line can be matched to the desired analysis accuracy and calculation time. What is necessary is just to set the division | segmentation number of a path | pass.

  Furthermore, by configuring the incident neutron flux distribution to be input based on an arbitrary function formula, it is possible to select a function formula that matches the actual neutron flux distribution and perform a highly accurate analysis.

  Further, by inputting the function expression, the storage capacity for storing the number of input data and the calculation progress can be greatly reduced.

  The best mode for carrying out the present invention will be described below with reference to the drawings.

  FIG. 2 shows a plan view of a core model 100 that models the core of a pressurized water reactor (PWR) to be analyzed in the present embodiment.

  The core contains the fuel assemblies and control rods (control rod clusters) in a nuclear reactor (pressure) vessel. A control rod cluster guide tube, a control rod drive unit, etc. are arranged in the upper part of the reactor core. The core baffle, heat shield, etc. are arranged.

  The core model 100 shown in FIG. 2 is formed from a plurality of fuel assembly models 10... That model fuel assemblies. For example, a 193 fuel assembly model having a substantially square shape in plan view. When 10 is collected in a circular shape, a 4-loop (four steam generators) core model 100 as shown in FIG. 2 is obtained.

  Here, the number of loops differs depending on the output of the nuclear reactor to be analyzed, and there are a 2-loop type in which 121 fuel assembly models 100 are inserted and a 3-loop type in which 157 bodies are inserted.

  In addition, the fuel assembly models 10,... Are not all loaded into the core at a time, but for example, the loading time is shifted three times so that the output does not increase at a time. In addition, the new fuel assembly model 10 is arranged on the outside, and the output distribution is flattened so that only the output at the center part away from the outer boundary is not increased.

  A fuel assembly model 10 to be loaded into such a core model 100 is formed as shown in FIG. 3 by loading a plurality of fuel rod models 1 obtained by modeling fuel rods as shown in FIG.

  Here, an example of the configuration of the fuel assembly will be described. 264 fuel rods forming a 17 × 17 square array, one in-core instrumentation guide thimble, 24 control rod guide thimbles, and support It is composed of nine grids and upper and lower nozzles.

  A fuel assembly model 10 is obtained by modeling elements that influence the analysis result among the elements constituting the fuel assembly.

  As another type of fuel assembly, there is a 14 × 14 type.

  On the other hand, the fuel rods loaded in the fuel assembly include, for example, low-concentrated uranium dioxide sintered pellets inserted into a Zircaloy-4 cladding tube, a stainless steel spring inserted in the upper part, and helium pressurized and sealed at both ends. -It is formed into a sealed structure by welding end plugs.

  In addition, fuel rods are subject to excessive stress on the cladding and end plug welds due to fission product gas released from the sintered pellets, differences in thermal expansion between the cladding and sintered pellets, or changes in fuel density associated with combustion. In order not to be added, an appropriate gap is provided between the sintered pellet and the cladding tube.

  As shown in FIG. 4 (a), this fuel rod has a cylindrical fuel region 1a filled with sintered pellets of low enriched uranium dioxide and an annular modeled model of a Zircaloy-4 cladding tube on the outer periphery thereof. The fuel rod model 1 is formed by a cladding tube region 1c and a water region 1b as a moderator region on the outer periphery thereof.

  In this embodiment, the fuel rod model 1 is analyzed as a three-dimensional solid cell.

  Here, a plane as a two-dimensional plane of the fuel rod model 1 is shown in FIG. 4 (b), and one of the fuel assembly model 10 in which the planes of the plurality of fuel rod models 1,. FIG. 3 shows the part.

  FIG. 3 shows a plurality of straight paths 2,... As tracks of neutrons passing through the plane. These straight paths 2,... Are tracks of straight neutrons drawn in parallel at a predetermined interval with an azimuth angle ψ, and the position is indicated by k.

  And the state which extended | stretched this linear path 2 to the axial direction of the fuel rod model 1 which is a direction orthogonal to a plane, and produced | generated the surface areas 31-33 was shown to Fig.1 (a).

  In FIG. 1 (a), only the fuel region 1a and the water region 1b of the fuel rod model 1 are shown for simplicity of explanation, and division in each region is not considered.

  The surface regions 31 to 33 are rectangular and form a surface region 31 that crosses the water region 1b, a surface region 32 that crosses the fuel region 1a, and a surface region 33 that crosses the opposite water region 1b.

  Here, the neutron flux passing through the surface region 31 will be described with reference to FIG. This neutron flux is the ratio of neutrons that pass through a certain point of the core in a certain direction, expressed as V × n by the velocity V (energy E) in the unit volume and the neutron density n in the certain direction. Can do.

  Further, the surface region 31 has a rectangular shape with a width w and a height h, and each side thereof is defined as a side side 3a, 3c, a bottom side 3b, and an upper side 3d.

  When the neutron flux passes through the surface region 31 at an elevation angle θ with respect to the base 3b, the neutron flux φin enters from the one side 3a, and the neutron flux φbottom also enters the surface region 31 from the base 3b. Is done. Here, the neutron flux φbottom is a code that specifically designates the incident position of the incident neutron flux.

  Further, the neutron flux φout is emitted from the other side 3c, and the neutron flux φtop is also emitted from the surface region 31 from the upper side 3d. This neutron flux φtop is also a code that specifically specifies the emission position of the emitted neutron flux.

  The relationship between the incident neutron fluxes φin and φbottom and the emitted neutron fluxes φout and φtop does not indicate individual tracks of neutrons.

  Then, paying attention to the neutron balance in the surface region 31, neutron fluxes φout and φtop emitted through the surface region 31 are calculated from the incident neutron fluxes φin and φbottom by the Characteristics method.

  In this Characteristics method, the Boltzmann equation that gives an accurate prediction of the transport phenomenon is continuously solved on the linear path 2, and the work is repeated on the multiple linear paths 2,. This is an analysis method for calculating the neutron flux φ to be finally obtained.

  In FIG. 1A, one straight path 2 is illustrated in the system divided by the fuel rod model 1, and Boltzmann is applied to three surface areas 31 to 33 continuous on the straight path 2. The equation will be solved.

  A relational expression of the neutron flux at the azimuth angle Ψ and the elevation angle θ of the surface region 31 at the position k is as follows.

In this equation, φ _p is the surface neutron flux φ (p, θ, Ψ) indicated by the average neutron flux of the fuel rod model 1, Q is the neutron source of the fuel rod model 1, and Σ is the all-macro disconnection of the fuel rod model 1. Indicates area. This total macro cross section is the probability that a neutron interacts with a nucleus in the material during a unit length flight through the material.

  And this plane neutron flux (phi) (p, (theta), (psi)) has shown the magnitude | size of the neutron flux of the surface area | region p (31).

  Subsequently, with respect to the surface region 32 adjacent to the surface region 31, the neutron flux φout emitted from the surface region 31 is used as the neutron flux φin incident on the surface region 32, and the surface neutron flux φ (p, θ, Ψ) is calculated and this calculation is repeated on the straight path 2 at position k.

  Although not shown here, the neutron flux φtop emitted from the upper side 3d of the surface region 31 becomes the neutron flux φbottom incident from the bottom of the surface region in contact with the upper side 3d of the surface region 31.

  Further, the other positions k-1, k + 1,... Are calculated in the same manner, and are integrated in the space of the fuel rod model 1 based on the results, whereby the elevation angle θ of the fuel rod model 1 is obtained. Then, the angle neutron flux φ (θ, Ψ) of the azimuth angle Ψ is calculated.

  This calculation is also performed for the remaining elevation angle θ and azimuth angle ψ, and the three-dimensional neutrons of the three-dimensional fuel rod model 1 are integrated by the elevation angle θ and azimuth angle ψ based on the results. The bundle φ is calculated.

  The number of straight paths 2, the number of elevation angles θ, and the number of azimuth angles Ψ can be arbitrarily set according to the desired calculation accuracy and time.

  Next, the flow of processing of the three-dimensional core analysis program of this embodiment will be described.

  A computer functioning by this three-dimensional core analysis program includes an input device such as a keyboard and a mouse, and a hard disk, ROM, RAM, etc. for storing programs and various data, and temporarily storing input values and calculation values. Mainly composed of a storage device, a central processing unit (CPU) that performs calculations based on values or instructions read from the storage device or input from the input device, and an output device such as a monitor or printer that outputs the calculation results Has been.

  Then, the input device and the storage device are caused to function by the input means of the three-dimensional core analysis program, and the output device is caused to function by the output means.

  In addition, nuclear data files required for core analysis such as reactor constants and group constants can be obtained from the Nuclear Data Center of the Japan Atomic Energy Research Institute, and these data and processed data are stored in a storage device. It can be read out as necessary.

  FIG. 5 is an example of a flowchart showing the overall processing flow of the three-dimensional core analysis program of the present embodiment.

  First, in step S1, analysis conditions such as initial values and shape data are read from the input by the input device and the data stored in the storage device.

  The analysis conditions include each shape data of the fuel rod models 1,... As a three-dimensional cell, data for setting the straight path 2 and the plane region p, analysis targets such as the fuel assembly model 10 and the core model 100. Model overall shape data, nuclear reaction cross section data of each fuel rod model 1,... Boundary conditions, initial value of 3D neutron flux φ of each fuel rod model 1,. Enter the function formula format of the incident neutron flux distribution.

  Subsequently, in step S2, straight line path data necessary for generating a surface area is generated, and combined with the height of the surface area input in step S1, a surface area is generated in a later step (surface area). Generating means).

  This linear path 2 is set at a plurality of positions k so as to cover the analysis range as shown in FIG. 3 with respect to the plane of the fuel rod models 1. The straight path 2 is set in a plurality of directions by changing the azimuth angle Ψ.

  The number of the straight lines 2 to be drawn (position k) and how many times the azimuth angle Ψ is set are input in step S1.

  The data of the straight path 2 set in the plane in this way is temporarily stored in a RAM or the like, and the width w of the straight path 2 and the height h of the surface area are shown in FIG. Plane areas 31 to 33 are generated for each of the straight paths 2,.

  Here, the following calculated values are also appropriately stored in a storage device such as a RAM and read when necessary.

  In this three-dimensional core analysis program, the energy region of neutrons is divided into a plurality of parts, and analysis is performed for each divided energy section. Therefore, in step S3, this energy group is designated.

  In step S4, the neutron source Q generated in the fuel rod model 1 is calculated from the neutron flux distribution.

  That is, in each fuel rod model 1,... And each energy group, the initial value or the latest three-dimensional neutron flux φ calculated in the convergence process and the nuclear reaction cross section, Then, neutrons and neutron flux generated by scattering of the nuclei are obtained, and a neutron source Q is obtained by adding them in each fuel rod model 1 for each energy group.

  Subsequently, a neutron flux distribution is calculated from the neutron source Q by a method described later (step S5), and the process is repeated until the desired neutron flux value converges (step S6). Here, “neutron flux converges” refers to a state in which the neutron flux obtained previously and the value of the neutron flux obtained based on the value are within the set error range.

  Since the calculation for obtaining the neutron flux distribution is performed for all of the divided energy groups, the same calculation is repeated if an uncalculated energy group remains (step S7).

  In step S8, the multiplication factor is calculated from the neutron flux calculated above.

  The multiplication factor K is a multiplication factor relating to nuclear fission, and is represented by K = (number of neutrons generated per unit time) / (number of neutron annihilation per unit time due to absorption and leakage).

  The multiplication factor K is critical when K = 1, is supercritical when K> 1 and fission increases with generations to cause chain reaction to diverge, and when K <1, the chain reaction is It is a subcritical state to stop.

  Then, the multiplication factor K when it is assumed that the fuel rod models 1 are arranged infinitely is called an infinite multiplication factor K∞.

  The calculation is repeated until the multiplication factor calculated in this way converges (step S9). Here, “the multiplication factor converges” refers to a state in which the multiplication factor obtained previously and the multiplication factor obtained based on the value are within the set error range.

  When the calculation is completed, a calculation result such as a multiplication factor is output to the output device in step S10.

  Next, the calculation of the neutron flux distribution in step S5 of FIG. 5 will be described in detail with reference to the flowchart of FIG.

  First, an energy group is specified in step S3 of FIG. 5, and the neutron source Q of each fuel rod model 1,... Is calculated in step S4.

  Then, the elevation angle θ of the incident neutron fluxes φin and φbottom is read in step S11 of FIG. 6, and the azimuth angle Ψ of the straight path 2 is read in step S12.

  Subsequently, in step S13, the neutron fluxes φin and φbottom, which are initial values at the boundary of the analysis region, are input when starting the iterative calculation of the neutron flux distribution calculation. In this input, the value input from the input device in step S1 may be stored in the storage device and read from the storage device at the start of calculation. Moreover, you may read the value memorize | stored in the memory | storage device beforehand as a database.

  Then, the surface neutron flux φ (p, θ, Ψ) of the surface region p is calculated from the neutron fluxes φin and φbottom input as initial values (step S14).

  Specifically, first, data of the width w and the height h of the straight path 2 is read from the storage device, and the surface areas 31 to 33 are generated.

  Further, as shown in FIG. 1, the surface neutron flux φ (p, θ, Ψ) of the surface regions 31 to 33 at the azimuth angle Ψ and the elevation angle θ of the linear path 2 at the position k 1 is calculated. At this time, the neutron flux incident on the surface region p can be expressed based on the functional expression input in step S1.

  Subsequently, the surface neutron flux φ (p, θ, Ψ) is similarly applied to the straight paths 2,... At positions k + 1, k−1,. calculate.

  The calculation of the surface neutron flux φ (p, θ, Ψ) is repeated until the neutron fluxes φin and φbottom incident on the surface region p converge (step S15).

  The calculation for calculating the plane neutron flux φ (p, θ, ψ) of the elevation angle θ and the azimuth angle ψ is performed for all the plane regions p in the space of the fuel rod model 1. If the analysis target is symmetrical, the analysis can be performed only on half (180 degrees) and ¼ (90 degrees) of the analysis target and developed into the entire system.

  Then, in step S16, these calculation results are integrated in the space of each fuel rod model 1,..., And the angle neutron flux φ (θ, Ψ) at the elevation angle θ of the azimuth angle Ψ is determined as the angle neutron calculation means. Calculate as

  Thereafter, by changing the value of the azimuth angle Ψ, the angular neutron flux φ (θ, Ψ) is calculated for all the azimuth angles Ψ set as the analysis conditions (step S17). Subsequently, the angle neutron flux φ (θ, Ψ) is calculated for all the elevation angles θ set as the analysis conditions by changing the value of the elevation angle θ (step S18).

  Then, the angle neutron flux φ (θ, Ψ) calculated for all azimuth angles Ψ and elevation angles θ is angularly integrated in the space of each fuel rod model 1,. The three-dimensional neutron flux φ is calculated as neutron calculation means (step S19).

  The calculation of the three-dimensional neutron flux φ is performed until convergence (step S6 in FIG. 5), and after convergence, the calculation is performed for the next energy group (step S7 in FIG. 5).

  Hereinafter, Example 1 of the above-described embodiment will be described. The description of the same or equivalent parts as those described in the above embodiment will be given the same reference numerals.

  In Example 1, an analysis example performed by causing a computer to function by the three-dimensional core analysis program of the above embodiment will be described.

  First, in Example 1, assuming a case where a 17 × 17 type fuel assembly is loaded into the core of a pressurized water reactor (PWR), a fuel rod model 1 filled with fuel having a uranium enrichment of 4.8% is assumed. Analysis was performed.

  Here, the fuel rod model 1 has a height of 5 cm, a diameter of the fuel region 1a of 0.82 cm, and each side of the water region 1b of 1.26 cm × 1.26 cm.

  Further, the periphery of the fuel rod model 1 was a completely reflective boundary, and the energy division was 70 groups.

  Further, as shown in FIG. 4 (b), the fuel rod model 1 includes a dividing line 11, which divides the fuel region 1a into five in addition to the boundary lines of the fuel region 1a, the cladding tube region 1c, and the water region 1b. And the dividing line 12 which divides the water area | region 1b into 2 was added as a dividing line.

  When the number of divisions of the fuel rod model 1 is increased in this way, the number of division lines that the straight path 2 crosses increases, and the number of divisions of the straight path 2 and the number of surface areas on the straight path 2 increase.

  Further, in order to verify the analysis result of the three-dimensional core analysis program of the present invention, a two-dimensional analysis was performed as a reference solution.

  From the results of Table 1, it can be seen that when the number of divisions of the fuel region 1a is five, the infinite multiplication factor tends to be higher than the two-dimensional analysis result as the axial height of the fuel rod model 1 decreases. .

  On the other hand, when the number of divisions of the fuel region 1a is increased to 10 divisions, the difference from the two-dimensional analysis result is reduced, so that the neutron flux is averaged at the axial boundary surface. It can be considered that the calculation result of the distribution is affected.

  In the first embodiment, as shown in FIG. 7A, the input values F1 of the neutron fluxes φin and φbottom incident on the side 3a and the bottom 3b of the surface region 31 are set as constants that are average values in the respective ranges. The neutron flux φout emitted from the surface region 31 is defined as the neutron flux φin that is incident on the adjacent surface region 32, and the neutron flux that is incident on the surface region (not shown) that is adjacent to the neutron flux φtop emitted from the surface region 31. φbottom.

On the other hand, the neutron flux φin and φbottom incident on the surface region 31 may be a quadratic function (ax ² + bx + c) or a linear function (dx +) as shown by the input function F2 in FIG. e) and so on.

  In this case, the neutron flux φin and φbottom incident on the surface region 32 adjacent to the surface region 31 and the surface region (not shown) on the upper surface can also be specified by the input function F2 and the like. The error due to can be reduced.

  Also, by changing the function coefficients (a to c and d, e) for each surface area, it is possible to perform a highly accurate analysis reflecting the position and characteristics of the surface area.

  That is, whether the incident neutron flux φin, φbottom is expressed by a constant or a function can be arbitrarily set depending on the calculation processing capacity, storage capacity, calculation time, and desired calculation accuracy of the computer.

  Other configurations and functions and effects are substantially the same as those in the above-described embodiment, and thus description thereof is omitted.

  Hereinafter, Example 2 of the above-described embodiment will be described. The description of the same or equivalent parts as those described in the above embodiment will be given the same reference numerals.

  In the second embodiment, as shown in FIG. 8A, an analysis example in which the fuel rod model 5 whose cross section changes in the axial direction is analyzed as a three-dimensional cell will be described.

  The fuel rod model 5 is a three-dimensional cell in which a reflector 52 having a height of 10 cm is disposed under a fuel body 51 having a height of 25 cm.

  The fuel body 51 includes a fuel region 51a having a diameter of 0.82 cm, a cladding tube region 51c having a diameter of 0.95 cm, and a water region 51b having a diameter of 1.26 cm × 1.26 cm. Further, the reflectors 52 are all water regions 52a.

  Further, the periphery of the fuel rod model 5 is a completely reflective boundary, the energy division is 70 groups, and the fuel region is divided into 10 divisions.

  In the analysis result of Example 2, the infinite multiplication factor of the entire analysis target was 1.2329, and the difference from the three-dimensional analysis result by the multi-group Monte Carlo method (GMVP, Japan Atomic Energy Research Institute) was 0.0014. FIG. 8B shows the axial distribution of relative neutron flux for neutron flux in a low energy region.

  From this analysis result, the difference from the three-dimensional analysis result of the infinite multiplication factor was slightly large, but the axial distribution of the neutron flux was sufficiently reproduced as shown in FIG. It is clear that the core analysis program is suitable for three-dimensional analysis.

  Other configurations and functions and effects are substantially the same as those of the above-described embodiment or Example 1, and thus description thereof is omitted.

  The preferred embodiments and examples of the present invention have been described in detail with reference to the drawings, but the specific configuration is not limited to the embodiments and examples, and does not depart from the gist of the present invention. A degree of design change is included in the present invention.

  For example, in this embodiment, the fuel rod model 1 is analyzed as a three-dimensional cell. However, the present invention is not limited to this, and the fuel assembly model 10 may be a three-dimensional cell, or a plan view of the core model 100 may be arbitrarily set. A rectangular parallelepiped having a two-dimensional plane formed by dividing into a mesh shape by the number of can be made into a three-dimensional cell.

  Further, the number of energy group divisions is not limited to 70 groups.

  Furthermore, although the number of divisions of the fuel region 1a has been described as being divided into five and ten, it is not limited to this.

  Moreover, although the detailed flowchart was shown in FIG.5, 6 shown in the said embodiment, it is not limited to this.

  Furthermore, in the above embodiment, the core of the pressurized water reactor (PWR) has been described. However, the present invention is not limited to this, and the present invention is applied to other core analysis such as a boiling water reactor, a gas cooling reactor, and the like. You can also.

BRIEF DESCRIPTION OF THE DRAWINGS It is a figure which shows the structure of the three-dimensional cell used with the three-dimensional core analysis method of the best embodiment of this invention, (a) is explanatory drawing which showed the linear path | pass and the surface area | region in the perspective view of a fuel rod model, (B) is explanatory drawing explaining the neutron flux which injects into a surface area | region, and the neutron flux inject | emitted from there. It is the top view which showed the structure of the core model. It is the fragmentary top view which showed the structure of the fuel assembly model. It is a figure explaining the detailed structure of a fuel rod model, (a) is a perspective view, (b) is a top view of the fuel rod model which added the dividing line. It is a flowchart which shows the flow of the whole process of the three-dimensional core analysis program of the best embodiment of this invention. It is a flowchart explaining the flow of a process of calculation of neutron flux distribution. It is a figure explaining the input value of the neutron flux which injects into a surface area, (a) is a constant and (b) shows a quadratic function. (A) is the perspective view explaining the structure of the fuel rod model of Example 2, (b) is the figure which showed the axial direction distribution of the analysis result.

Explanation of symbols

100 Core model 10 Fuel assembly model 1 Fuel rod model (three-dimensional cell)
2 Straight path 31-33 Surface area F1 Input value (constant)
F2 input function (function formula)
5 Fuel rod model (three-dimensional cell)
φ Three-dimensional neutron flux φ (p, θ, Ψ) Plane neutron flux φ (θ, Ψ) Angle neutron flux Ψ Azimuth angle θ Elevation angle

Claims (2)

A three-dimensional core analysis method for analyzing neutron behavior in a three-dimensional cell in a system in which a core including a plurality of fuel rods is divided into a plurality of three-dimensional solid cells.
A neutron track passing through the surface at a certain azimuth angle Ψ with respect to the two-dimensional surface of the three-dimensional cell is defined by a plurality of linear paths,
Generating a quadrangular surface region with the upper and lower sides being a line segment divided by the point where the straight line path intersects the dividing line arbitrarily dividing the top surface and the bottom surface of the solid cell to be the two-dimensional surface ;
A neutron flux incident at an elevation angle θ by a distribution represented by a linear function or a quadratic function with respect to the base and side of the surface area, and a neutron flux emitted from the top and side sides, respectively , of the surface area While calculating the surface neutron flux, calculating the angle neutron flux of the three-dimensional cell by calculating the surface neutron flux for all of the surface areas on the plurality of linear paths described above,
The same calculation is performed for a plurality of the azimuth angles Ψ and the elevation angle θ to calculate all the angular neutron fluxes, and the angular integration is performed to calculate the three-dimensional neutron flux of the three-dimensional cell. analysis method. A three-dimensional core analysis program for causing a computer to function to analyze the behavior of neutrons in a three-dimensional cell in a system in which a core including a plurality of fuel rods is divided into a plurality of three-dimensional solid cells.
An input means for inputting analysis conditions;
Based on the analysis conditions, a track of neutrons passing through the two-dimensional plane of the three-dimensional cell at a certain azimuth angle Ψ is defined by a plurality of parallel straight paths, and the azimuth angle Ψ is changed in the same manner. Generate a straight path for each angle Ψ,
A surface that generates a quadrangular surface region p having a top and bottom sides of a line segment that is divided at a point where a dividing line that arbitrarily divides the top surface and bottom surface of the three-dimensional cell that becomes the two-dimensional surface and the straight line path intersect. Region generation means;
The surface region from the neutron flux incident at an elevation angle θ by the distribution expressed by a linear function or a quadratic function with respect to the base and side of the surface region p, and the neutron flux emitted from the upper side and the side, respectively. The surface neutron flux φ (p, θ, Ψ) in the inner space is calculated for all straight paths, and the surface neutron flux φ (p, θ, Ψ) in the surface region p included in the space of the three-dimensional cell is calculated in space. The angle neutron calculating means for calculating the angle neutron flux φ (θ, Ψ) of the three-dimensional cell toward the elevation angle θ and the azimuth angle Ψ by integrating within
Neutron flux calculating means for calculating the three-dimensional neutron flux φ in the three-dimensional cell by integrating the angular neutron flux φ (θ, Ψ) with an angle;
A three-dimensional core analysis program for causing an output means for outputting an analysis result based on the three-dimensional neutron flux φ to function .

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