Method for calculation, nuclear core analyzer, and program

Abstract

To provide a method for making a burning calculation in a nuclear core analysis more rapid.SOLUTION: The method for calculation evaluates nuclear characteristics by the steps of: determining a small group which can maintain the number density calculated by a neutron transfer calculation and a burning calculation in a minute group; calculating the number density of each step up to an evaluation target step in the small group; and performing a neutron transfer calculation in the minute group by using the calculated number density as input for the evaluation target step.SELECTED DRAWING: Figure 1

Description

Technical Field

[0001] The present disclosure relates to a calculation method, a core analysis device, and a program.

Background Art

[0002] In core analysis, the process in which the composition of the in-core fuel changes during operation due to nuclear fission, neutron absorption, etc. is called combustion, and the calculation for tracing the change in the number density of nuclides due to combustion is called a combustion calculation. In a core calculation code, a combustion calculation is performed using the neutron flux distribution obtained by neutron transport calculation or diffusion calculation, and after updating the number density of nuclides in the next step, the nuclear characteristics in that state are evaluated. In order to evaluate the nuclear characteristics at any step from the initial loading core where the in-core fuel is fresh fuel to the end of the cycle and in subsequent cycles succeeding that in-core fuel, combustion calculations up to the evaluation target step are necessary. Neutron transport calculations often take time, and the calculation time for the entire combustion calculation up to the evaluation target step also often becomes long. Especially in the design of a fast reactor core, since neutron transport calculations are carried out under detailed conditions with 70 energy groups, combustion calculations may take several days to several tens of days, and until the calculations are completed, the nuclear characteristics of the evaluation target step cannot be evaluated.

[0003] Patent Document 1 discloses a method for shortening the calculation time by setting simple parameters when performing neutron transport calculations based on a large number of energy groups in calculating the nuclear constants of the fuel assemblies loaded in the core, and shortening the calculation time by performing neutron transport calculations based on a small number of energy groups when performing neutron transport calculations by setting detailed parameters after the calculations.

Prior Art Documents

Patent Documents

[0004]

Patent Document 1

Summary of the Invention

Problems to be Solved by the Invention

[0005] Provided is a method for accelerating combustion calculations in core analysis.

[0006] The present disclosure provides a calculation method, a core analysis device, and a program capable of solving the above problems.

Means for Solving the Problems

[0007] The calculation method according to the present disclosure includes: determining a first number density of a predetermined nuclide obtained by performing a combustion calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics, and a second number density of the nuclide obtained by performing a combustion calculation based on a neutron flux distribution calculated by neutron transport calculation using a smaller number of minority groups than the detailed number of neutron energy groups, and determining the number of minority groups such that the difference therebetween is within a predetermined allowable range; performing, in time series, the neutron transport calculation using the number of minority groups and the combustion calculation based on the neutron flux distribution calculated by the neutron transport calculation at each evaluation target time point at which evaluation of the nuclear characteristics is required and at each predetermined intermediate time point up to the evaluation target time point, thereby calculating the number density at each evaluation target time point and each intermediate time point; and calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input.

[0008] The core analysis device according to the present disclosure includes means for determining a minority group number such that the difference between a first number density of a predetermined nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a minority group number smaller than the detailed number of neutron energy groups is within a predetermined allowable range; means for calculating the number density at each of the evaluation target time point and the intermediate time points by performing the neutron transport calculation using the minority group number and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation in time series for each of the evaluation target time point when evaluation of the nuclear characteristics is required and a predetermined intermediate time point up to the evaluation target time point; and means for calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input.

[0009] The program according to the present disclosure causes a computer to execute a step of determining a minority group number such that the difference between a first number density of a predetermined nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a minority group number smaller than the detailed number of neutron energy groups is within a predetermined allowable range; a step of calculating the number density at each of the evaluation target time point and the intermediate time points by performing the neutron transport calculation using the minority group number and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation in time series for each of the evaluation target time point when evaluation of the nuclear characteristics is required and a predetermined intermediate time point up to the evaluation target time point; and a step of calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input.

Advantages of the Invention

[0010] According to the calculation method, core analysis device, and program of the present disclosure, the combustion calculation in core analysis can be accelerated.

Brief Description of the Drawings

[0011]

Figure 1

Figure 2

Figure 3

Figure 4

Figure 5

Figure 6

Modes for Carrying Out the Invention

[0012] <Embodiment> Hereinafter, the core analysis device of the present disclosure will be described with reference to FIGS. 1 to 6. (Configuration) FIG. 1 is a block diagram showing an example of a core analysis device according to an embodiment. The core analysis device 10 executes the combustion calculation in core analysis in a shorter time than before. The core analysis device 10 includes an input reception unit 11, a group number determination unit 12, a number density calculation unit 13, a nuclear characteristic calculation unit 14, an output unit 15, and a storage unit 16.

[0013] The input reception unit 11 receives information, instructions, etc. input using input devices such as keyboards, mice, touch panels, buttons, etc. For example, the input reception unit 11 receives inputs such as calculation conditions necessary for core analysis and an execution instruction for core analysis. The input reception unit 11 records the received information in the storage unit 16 or outputs it to the group number determination unit 12, the number density calculation unit 13, and the nuclear characteristic calculation unit 14.

[0014] The group number determination unit 12 determines the number of energy groups used in neutron transport calculations. Since the number density of nuclides depends on the reaction rate, the number of energy groups is determined so that the reaction rate calculated by neutron transport calculation with a detailed number of energy groups can be maintained for nuclear reactions of the main substances for evaluating nuclear characteristics. In other words, the number of energy groups is determined so that the calculation accuracy of the number density equivalent to that obtained by the burnup calculation based on neutron transport calculation with a detailed number of energy groups is achieved. The number of energy groups determined by the group number determination unit 12 is a value smaller than the detailed number of energy groups. For example, in the case of a fast reactor, the detailed number of energy groups is 70 groups, and the number of energy groups determined by the group number determination unit 12 is 20 to 30 groups. Also, the group number determination unit 12 may determine an even smaller number of groups according to the required calculation accuracy. The detailed number of energy groups is called the detailed group, and the small number of energy groups determined by the group number determination unit 12 is called the small group.

[0015] Referring to FIG. 2, a method for determining the number of energy groups will be described. The vertical axis of the graph in FIG. 2 represents the reaction rate, and the horizontal axis represents the neutron energy (eV). Graph L1 represents the reaction rate by neutron energy of a nuclear reaction (for example, U-238 capture) of a certain nuclide existing in the core to be analyzed. The reaction rate by neutron energy for each nuclear reaction of a nuclide (hereinafter sometimes simply referred to as "for each nuclide") can be obtained by performing neutron transport calculations. Graph L1 is the calculation result obtained by performing neutron transport calculations based on the detailed groups. When creating the few-group structure, the group number determination unit 12 confirms the reaction rate for each nuclear reaction of the nuclides in the core system to be analyzed based on Graph L1, and calculates the neutron energy band that can be coarsened as the group structure. For example, the group number determination unit 12 determines the range of neutron energy within which the change in the reaction rate is within a predetermined range as the range where the energy groups can be coarsened. In the case of the example in FIG. 2, the group number determination unit 12 determines, for example, ranges 21 and 23 as the ranges where the energy groups can be coarsened. For example, the group number determination unit 12 reduces the number of energy groups in range 21 corresponding to a dozen or so groups out of the 70 groups when divided into detailed groups from a dozen or so groups to several groups. Similarly, the group number determination unit 12 reduces the number of energy groups in range 23 corresponding to a dozen or so groups when divided into detailed groups to several groups. The group number determination unit 12 may maintain the number of groups for range 22 or slightly reduce the number of groups. The group number determination unit 12 holds data on the cross-section area for each nuclide in the few-group. The group number determination unit 12 performs neutron transport calculations based on the few-group using the cross-section area of the few-group, and calculates the reaction rate by neutron energy for each nuclide. The reaction rate by neutron energy of the same nuclide calculated in the few-group is shown in graph L2 of FIG. 2. The group number determination unit 12 evaluates the difference between the reaction rate L1 calculated in the detailed group and the reaction rate L2 calculated in the few-group for each neutron energy band of a predetermined width, for example. When the differences between L1 and L2 for each predetermined width are all within the allowable range, it is determined that the number of energy groups for the nuclear reaction of the nuclide has been determined. The group number determination unit 12 performs the same evaluation for other nuclides, and searches for the number of energy groups that satisfy the conditions for the nuclear reactions of all nuclides in the core system (that is, the difference between the reaction rate calculated in the detailed group and the reaction rate calculated in the assumed few-group is within the allowable range).The group number determination unit 12 may search for the minimum number of energy groups that satisfy the conditions. Alternatively, if an energy group number that satisfies the conditions is found for all nuclides between 20 and 30 groups, that energy group number may be determined as the number of groups for the minority groups. Further, the group number determination unit 12 may determine the minimum number of groups with an even smaller number of energy groups than the minority groups (20 to 30 groups) based on a second condition with relaxed conditions within the allowable range. The method for determining the minimum number of groups is the same as the method for determining the number of minority groups described above, except that the allowable range of the difference from the reaction rate calculated in the detailed groups is widened. Note that the determination of whether the difference in the reaction rate between the graph L1 obtained from the transport calculation in the detailed groups and the graph L2 obtained from the transport calculation in the minority groups is within the allowable range for the settings in the range 21 to 23 may be performed by the user.

[0016] The number density calculation unit 13 uses a core calculation code to perform neutron transport calculations with the number of energy groups set to the minority groups, and performs burnup calculations based on the neutron measurement distribution of the calculation results. This calculation is called a burnup calculation based on the minority groups. The number density calculation unit 13 executes the burnup calculation based on the minority groups in time series for each predetermined calculation step from the beginning to the end of the evaluation target period.

[0017] The nuclear property calculation unit 14 uses a core calculation code to perform neutron transport calculations with the number of energy groups set to the detailed groups, and calculates nuclear properties (effective multiplication factor, reactivity, etc.). This calculation is called a nuclear property calculation based on the detailed groups. The nuclear property calculation unit 14 executes the nuclear property calculation based on the detailed groups only for the steps of the evaluation target.

[0018] The output unit 15 outputs the nuclear properties calculated in the burnup calculation to a display device, an electronic file, etc. The storage unit 16 stores the calculation conditions necessary for the burnup calculation (e.g., the fuel loading pattern, burnup distribution, evaluation target period, evaluation target steps, etc. for each cycle), the core calculation code, which is a program for performing neutron transport calculations and burnup calculations, the data during the calculation, and the calculation results.

[0019] Next, referring to FIG. 3, the evaluation target period and the evaluation target steps will be described. The evaluation target period is, for example, 10 cycles. One cycle indicates the period from the start or restart of the reactor core operation to the stop of the operation at the next regular inspection (for example, 13 months), and a step represents a predetermined period within one cycle. For example, when dividing one cycle into three steps, the start point of one cycle can be set as step 1, the midpoint of one cycle as step 2, and the end point of one cycle as step 3. When one cycle ends, some of the fuel loaded in the reactor core is removed, and other fuel is loaded instead to start the operation of the next cycle. For example, when it is desired to evaluate the last nuclear characteristics of the 10th cycle, the evaluation target period is 10 cycles, and the evaluation target step is step 3 of the 10 cycles.

[0020] When calculation conditions such as initial values (such as burnup) related to the state of the reactor core, the evaluation target period (for example, 10 cycles), the calculation step length (for example, the length between steps 1 to 3 of each cycle), and the fuel loading pattern for each cycle are given to the reactor core calculation code to instruct the execution of reactor core analysis, the reactor core state for each step within the evaluation target period is analyzed. In the conventional burnup calculation (referred to as the conventional method) in reactor core analysis, neutron transport calculations and burnup calculations based on a detailed group are performed for step 1 of one cycle, neutron transport calculations and burnup calculations based on a detailed group are performed for step 2 of one cycle, and so on, and such calculations are repeatedly executed until the end of the evaluation target period (step 3 of 10 cycles). Since neutron transport calculations based on a detailed group are performed every time at each step, the conventional method is likely to have a long calculation time. Also, in the conventional method, the number density and other nuclear characteristics (such as effective multiplication factor, reactivity, neutron flux distribution, etc.) are calculated for each step.

[0021] (Burnup Calculation 1) In contrast, in the present embodiment, the burnup calculation is divided into a process for calculating the number density for each nuclide and a process for calculating other nuclear characteristics. First, a burnup calculation based on a few groups is performed to calculate the number density for each step over the entire evaluation period. Then, the nuclear characteristic calculation based on detailed groups is performed only for the evaluation target step. At this time, the number density obtained as a result of the burnup calculation based on the few groups executed previously is used as input data for the neutron transport calculation in the nuclear characteristic calculation based on the detailed groups. In the case of the present embodiment, in the burnup calculation based on the few groups, since the neutron transport calculation using the few groups is executed, the calculation load is smaller than that of the conventional method, and the calculation can be completed in a relatively short time. Also, in the case of the present embodiment, since the few groups are determined so that a number density equivalent to that obtained by the neutron transport calculation and burnup calculation using the detailed groups is obtained, the number density at the evaluation target step can be calculated with accuracy comparable to that of the conventional method. Further, in the nuclear characteristic calculation based on the detailed groups, since only the neutron transport calculation independent of the burnup calculation is performed for the evaluation target step, the nuclear characteristics can be calculated in a shorter time than the conventional method.

[0022] The procedure of the burnup calculation according to the present embodiment in the case of the example of FIG. 3 will be specifically described. First, the number density calculation unit 13 calculates the number density of the nuclide to be analyzed in one step of one cycle by the number density calculation based on the few groups, updates the value of the number density of the nuclide to be analyzed with the calculated number density, and then calculates the number density in the second step of one cycle, and repeats this process step by step up to the third step of 10 cycles. Next, the nuclear characteristic calculation unit 14 uses the number density of the third step of 10 cycles calculated by the number density calculation unit 13 as input, and performs a neutron transport calculation based on detailed groups for the third step of 10 cycles (reflecting the loading pattern and burnup degree in the third step of 10 cycles) to calculate the nuclear characteristics with high accuracy. Thereby, the nuclear characteristics of the reactor core can be calculated in a shorter time and with the same degree of accuracy as compared with the conventional method.

[0023] (Burnup Calculation 2) Next, referring to FIG. 4, a method for more efficiently performing combustion calculations in a short time will be described for cases where the evaluation target period is as long as 10 cycles, or when all the fuel is aligned with new fuel at the start of one cycle (for example, at the start of operation of a new plant). As described above, every time one cycle ends, some of the fuel is replaced with other fuel during regular inspections. When all the fuel is aligned with new fuel at the start of one cycle, the fuel loaded into the reactor core is all new fuel in the first cycle. In the next two cycles, some of the fuel is removed and replaced with new fuel after one cycle ends, and the remaining fuel is used fuel (that has undergone one cycle of operation). In the next three cycles, some of the fuel is removed and replaced with new fuel after two cycles end, and the remaining fuel is used fuel (that has undergone one or two cycles of operation). By repeating such fuel replacement, for example, when it is desired to evaluate the nuclear characteristics at the last step of 10 cycles, the analysis results up to the first half cycle where the nuclear characteristics change for each cycle are not very useful. In such a case, for the first half of five cycles, neutron transport calculations are performed with the minimum number of groups (for example, around 10 groups) with an even further reduced number of energy groups, and for the second half of six cycles and later, neutron transport calculations are performed with a small number of groups (for example, 20 - 30 groups). According to this method, by calculating the first half of five cycles with the minimum number of groups, the calculation time can be further shortened compared to the combustion calculations described in FIG. 3. Also, for the second half, by performing neutron transport calculations with a small number of groups, the calculation accuracy of the number density calculation can be ensured.

[0024] The procedure of the combustion calculation according to the present embodiment in the case of the example of FIG. 4 will be specifically described. First, the number density calculation unit 13 calculates the number density of the nuclides to be analyzed step by step from step 1 of cycle 1 to step 3 of cycle 5 by neutron transport calculation and combustion calculation based on the minimum number of groups. Next, the number density calculation unit 13 continues with the number density of step 3 of cycle 5 and calculates the number density from step 1 of cycle 6 to step 3 of cycle 10 step by step by number density calculation based on the few groups. Next, the nuclear property calculation unit 14 executes neutron transport calculation based on the detailed group for step 3 of cycle 10 with the number density at step 3 of cycle 10 calculated by the number density calculation unit 13 as the input. As a result, the nuclear properties of the core can be calculated in a shorter time and with the same degree of accuracy compared to the conventional method. In addition, the nuclear properties of the core can be calculated in a shorter time than the combustion calculation described with reference to FIG. 3.

[0025] The following methods can be considered for the method of dividing the period of calculating the number density with the minimum number of groups and the period of calculating the number density with the few groups. (1) Divide the period from the initial operation to the end (life) of the plant into the first half and the second half in terms of time, and calculate the number density with the minimum number of groups in the first half and the few groups in the second half. (2) Calculate the number density with the minimum number of groups in the first half until the change in the nuclear properties of the core saturates, and calculate the number density with the few groups in the second half after the change in the nuclear properties of the core saturates. (3) Divide the steps up to the steps where the evaluation of the nuclear properties is required into the first half and the second half in terms of time, and calculate the number density with the minimum number of groups in the first half and the few groups in the second half. For example, when the operation of a total of 10 cycles is planned and the nuclear properties at step 3 of cycle 6 are to be evaluated, calculate the number density with the minimum number of groups from step 1 to step 3 and the few groups from step 4 to step 6. In the cases of (1) and (3), the timing for dividing into the first half and the second half can be set arbitrarily. For example, it may be divided so that the time lengths of the first half and the second half are 5:5, or 6:4, or 4:6.

[0026] (Operation) Next, the flow of the combustion calculation according to this embodiment will be described with reference to FIG. 5. FIG. 5 is a flowchart showing an example of the combustion calculation according to the embodiment. It is assumed that the storage unit 16 stores calculation conditions and the like necessary for the combustion calculation. First, the user sets the cycles and steps for evaluating the nuclear characteristics (step S1). For example, the user sets 10 cycles and 3 steps as the evaluation target steps. The input reception unit 11 acquires the input information and records it in the storage unit 16. Next, the user sets the period for calculating the number density in the few-group and the period for calculating the number density in the least few-group (step S2). For example, in the example of FIG. 3, the user sets the period for calculating the number density in the few-group as 1 cycle to 10 cycles and sets "Null" as the period for calculating the number density in the least few-group. For example, in the example of FIG. 4, the user sets the period for calculating the number density in the least few-group as 1 cycle to 5 cycles and the period for calculating the number density in the few-group as 6 cycles to 10 cycles. The input reception unit 11 acquires the input information and records it in the storage unit 16.

[0027] Next, the user instructs the determination of the number of groups. The input reception unit 11 receives this instruction and instructs the group number determination unit 12 to determine the energy group structure (how many neutron energies to divide into) of the few-group and the least few-group. The group number determination unit 12 determines the number of energy groups such that a reaction rate equivalent to the reaction rate obtained by the combustion calculation based on the detailed group is obtained (step S3). When calculating the number density in the least few-group, the allowable range for deviation from the reaction rate obtained by the combustion calculation based on the detailed group is widened, and a smaller number of energy groups than the few-group is determined. For example, in the case of a fast reactor with 70 detailed groups, the group number determination unit 12 determines the number of energy groups as 20 to 30 for the few-group and around 10 for the least few-group.

[0028] Next, the user instructs the start of the combustion calculation. The input reception unit 11 receives this instruction and instructs the number density calculation unit 13 to calculate the number density. The number density calculation unit 13 uses the core analysis code to perform number density calculations step by step up to the evaluation target step (step S4). For example, if it is set in step S2 to perform number density calculations in a few groups from 1 cycle to 10 cycles, the number density calculation unit 13 performs number density calculations in a few groups from step 1 of 1 cycle to step 3 of 10 cycles. For example, if it is set in step S2 to calculate in the fewest groups from 1 cycle to 5 cycles and perform number density calculations in a few groups from 6 cycles to 10 cycles, the number density calculation unit 13 performs number density calculations in the fewest groups from step 1 of 1 cycle to step 3 of 5 cycles and performs number density calculations in a few groups from step 1 of 6 cycles to step 3 of 10 cycles. The number density calculation unit 13 records the number density calculated for each step in the storage unit 16. Even if the operation is planned for a total of 10 cycles and the current evaluation target step is the nuclear characteristics at step 3 of 6 cycles, it may be configured to perform number density calculations in a few groups up to step 3 of 10 cycles and record the number density calculated for each step. By configuring it in this way, when it becomes necessary to evaluate the nuclear characteristics of step 3 of 10 cycles later, there is no need to execute the process of step S4 again, and by using the number density of step 3 of 10 cycles as the input and performing nuclear characteristic calculations for step 3 of 10 cycles, it is possible to evaluate the nuclear characteristics at the time of step 3 of 10 cycles.

[0029] When the number density calculation up to the evaluation target step is completed, the number density calculation unit 13 outputs the number density of the nuclides in the evaluation target step to the nuclear characteristic calculation unit 14 and instructs the execution of nuclear characteristic evaluation. The nuclear characteristic calculation unit 14 uses the core analysis code to perform neutron transport calculations based on detailed groups for the evaluation target step (step S5). Thereby, nuclear characteristics can be calculated accurately in a short time. Next, the output unit 15 outputs the nuclear characteristics in the evaluation target step to a display device or the like (step S6).

[0030] (Effect) As described above, according to the present embodiment, the combustion calculation in core analysis is divided into a process of calculating the number density and a process of calculating nuclear characteristics. In the process of calculating the number density, neutron transport calculation is performed with a small number of groups with a light calculation load, and in the process of calculating nuclear characteristics, neutron transport calculation based on detailed groups is performed only for the evaluation target step. Thereby, the combustion calculation in core analysis can be accelerated.

[0031] FIG. 6 is a diagram showing an example of the hardware configuration of the core analysis device. The computer 900 includes a CPU 901, a main storage device 902, an auxiliary storage device 903, an input / output interface 904, and a communication interface 905. The above-described core analysis device 10 is implemented in the computer 900. And each of the above-described functions is stored in the auxiliary storage device 903 in the form of a program. The CPU 901 reads the program from the auxiliary storage device 903 and expands it in the main storage device 902, and executes the above processing according to the program. Further, the CPU 901 secures a storage area in the main storage device 902 according to the program. Further, the CPU 901 secures a storage area in the auxiliary storage device 903 for storing data during processing according to the program. The core analysis device 10 may be configured by a plurality of computers 900. For example, the number density calculation unit 13 and the nuclear characteristic calculation unit 14 in FIG. 1 may be implemented in separate computers, or the number density calculation unit 13 and the nuclear characteristic calculation unit 14 may be implemented in a plurality of computers and configured to perform combustion calculation in core analysis in parallel with a plurality of computers.

[0032] A program for realizing all or part of the functions of the core analysis device 10 may be recorded on a computer-readable recording medium, and the program recorded on this recording medium may be read into a computer system and executed to perform the processing by each functional unit. The "computer system" referred to here shall include hardware such as an OS and peripheral devices. Also, the "computer system" shall include a homepage providing environment (or display environment) if the WWW system is being used. Further, the "computer-readable recording medium" refers to portable media such as CDs, DVDs, USBs, and storage devices such as hard disks built into a computer system. Also, when this program is distributed to the computer 900 via a communication line, the computer 900 that has received the distribution may expand the program in the main storage device 902 and execute the above processing. Also, the above program may be for realizing a part of the functions described above, and may further be capable of being realized in combination with a program already recorded in the computer system for the functions described above.

[0033] As described above, some embodiments according to the present disclosure have been described, but all of these embodiments are presented as examples and are not intended to limit the scope of the invention. These embodiments can be implemented in various other forms, and various omissions, replacements, and changes can be made without departing from the gist of the invention. These embodiments and their modifications are included in the scope and gist of the invention, and are also included in the invention described in the claims and the equivalent scope thereof.

[0034] <Supplementary Note> The calculation method, core analysis device, and program described in the embodiment are understood as follows, for example.

[0035] (1) The calculation method according to the first aspect includes: determining a minority group number such that the difference between a first number density of a predetermined nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance whose nuclear characteristics are to be evaluated and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a minority group number smaller than the detailed number of neutron energy groups is within a predetermined allowable range; calculating the number density for each of the evaluation target time point and the predetermined intermediate time points up to the evaluation target time point by performing, in time series, the neutron transport calculation using the minority group number and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation for each of the evaluation target time point and the predetermined intermediate time points up to the evaluation target time point; and calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input. Thereby, the burnup calculation in core analysis can be accelerated.

[0036] (2) The calculation method according to the second aspect is the calculation method according to (1), wherein in the step of calculating the number density, the period up to the evaluation target time point is divided into the first half and the second half, and for the first half, a neutron transport calculation using a group number even smaller than the minority group number and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation are performed, and for the second half, a neutron transport calculation using the minority group number and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation are performed. Thereby, the burnup calculation in core analysis can be further accelerated.

[0037] (3) The calculation method according to the third aspect is the calculation method according to (2), wherein when dividing the period up to the evaluation target time point into the first half and the second half, the period until the nuclear characteristics of the core are saturated is set as the first half, and the subsequent period is set as the second half. As a result, the combustion calculation in the core analysis can be further accelerated.

[0038] (4) The calculation method according to the fourth aspect is the calculation methods of (1) to (3), and in the step of determining the minority group number, the difference between the reaction rate calculated by the neutron transport calculation using the detailed neutron energy group number and the reaction rate calculated by the neutron transport calculation using the minority group number is within a predetermined allowable range, and the minority group number is determined. Thereby, an energy group number can be calculated such that the calculation accuracy of the number density does not decrease.

[0039] (5) The calculation method according to the fifth aspect is the calculation methods of (1) to (4), and in the step of calculating the number density, the number density up to a predetermined time point in the future from the evaluation target time point is calculated, and the calculated number density is recorded in a storage unit. For example, by calculating the number density over the entire range of the operation period predicted in advance, when it becomes necessary to evaluate the nuclear characteristics at a certain time point later, the neutron transport calculation based on the detailed group is performed using the number density at that time point as an input, so that the nuclear characteristics can be evaluated, and the combustion calculation in the core analysis can be further accelerated.

[0040] (6) The core analysis device according to the sixth aspect performs a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics, to obtain a first number density of a predetermined nuclide. A difference between the first number density and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a small number of groups less than the detailed number of neutron energy groups is within a predetermined allowable range. Means for determining the small number of groups, and the neutron transport calculation using the small number of groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation are executed in time series for each evaluation target time point at which evaluation of the nuclear characteristics is required and each predetermined intermediate time point up to the evaluation target time point, thereby calculating the number density for each of the evaluation target time point and the intermediate time point. Means for calculating, and for the evaluation target time point, means for calculating nuclear characteristics at the evaluation target time point by executing neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input.

[0041] (7) The program according to the seventh aspect causes a computer to perform a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance whose nuclear characteristics are to be evaluated, thereby obtaining a first number density of a predetermined nuclide, and a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a smaller number of groups than the detailed number of neutron energy groups, thereby obtaining a second number density of the nuclide, and determining the number of groups such that the difference between the first and second number densities is within a predetermined allowable range; performing the neutron transport calculation using the number of groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation for each time point of an evaluation target time point at which evaluation of the nuclear characteristics is required and each predetermined intermediate time point up to the evaluation target time point in time series, thereby calculating the number density for each of the evaluation target time point and the intermediate time point; and performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input for the evaluation target time point, thereby calculating the nuclear characteristics at the evaluation target time point.

Explanation of Signs

[0042] 10···Core analysis device 11···Input reception unit 12···Group number determination unit 13···Number density calculation unit 14···Nuclear characteristics calculation unit 15···Output unit 16···Storage unit 900···Computer 901···CPU 902···Main memory device 903···Auxiliary storage device 904···Input / output interface 905···Communication interface

Claims

1. A first number density of a predetermined nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics, and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a smaller number of groups than the detailed number of neutron energy groups, and determining the number of the smaller groups such that the difference therebetween is within a predetermined allowable range; Calculating the number density for each of the evaluation target time point and the intermediate time points by performing, in time series, the neutron transport calculation using the number of the smaller groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation, for each of the evaluation target time point at which evaluation of the nuclear characteristics is required and a predetermined intermediate time point up to the evaluation target time point; Calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input; A calculation method comprising the above steps.

2. In the step of calculating the number density, the period up to the evaluation target time point is divided into the first half and the second half. For the first half, a neutron transport calculation using a smaller number of groups than the number of the smaller groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation are performed. For the second half, a neutron transport calculation using the number of the smaller groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation are performed. The calculation method according to Claim 1.

3. When dividing the period up to the evaluation target time point into the first half and the second half, the period until the nuclear characteristics of the reactor core saturate is set as the first half, and the subsequent period is set as the second half. The calculation method according to Claim 2.

4. In the step of determining the number of the smaller groups, the number of the smaller groups is determined such that the difference between the reaction rate calculated by neutron transport calculation using the detailed number of neutron energy groups and the reaction rate calculated by neutron transport calculation using the number of the smaller groups is within a predetermined allowable range. The calculation method according to Claim 1 or Claim 2.

5. In the step of calculating the number density, the number density up to a predetermined time point in the future from the evaluation target time point is calculated, and the calculated number density is recorded in a storage unit. The calculation method according to Claim 1 or Claim 2.

6. Means for determining the number of minority groups such that the difference between a first number density of a predetermined nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a number of minority groups smaller than the detailed number of neutron energy groups is within a predetermined allowable range; Means for calculating the number density at each of the evaluation target time point and the intermediate time points by performing, in time series, the neutron transport calculation using the number of minority groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation at each of the evaluation target time point and predetermined intermediate time points up to the evaluation target time point at which evaluation of the nuclear characteristics is required; Means for calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input for the evaluation target time point; A core analysis device having the above.

7. On a computer, A step of determining the number of minority groups such that the difference between a first number density of a predetermined nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a predetermined detailed number of neutron energy groups for a substance for evaluating nuclear characteristics and a second number density of the nuclide obtained by performing a burnup calculation based on a neutron flux distribution calculated by neutron transport calculation using a number of minority groups smaller than the detailed number of neutron energy groups is within a predetermined allowable range; A step of calculating the number density at each of the evaluation target time point and the intermediate time points by performing, in time series, the neutron transport calculation using the number of minority groups and the burnup calculation based on the neutron flux distribution calculated by the neutron transport calculation at each of the evaluation target time point and predetermined intermediate time points up to the evaluation target time point at which evaluation of the nuclear characteristics is required; A step of calculating the nuclear characteristics at the evaluation target time point by performing a neutron transport calculation using the detailed number of neutron energy groups with the number density calculated for the evaluation target time point as an input for the evaluation target time point; A program for causing the above to be executed.

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