Radiation source strength evaluation method, shielding analysis method, radiation source strength evaluation device and program

Abstract

To provide a method for evaluating the intensity of a radiation source of a spent fuel including uncertainty.SOLUTION: A radiation source intensity evaluation method includes the steps of: evaluating the uncertainty of radiation source intensity resulting from nuclear data on the basis of the amount of generation of nuclides calculated by a calculation code for calculating disintegration and generation of the nuclides when the value of the nuclear data input to the calculation code is changed; calibrating the uncertainty of the radiation source intensity resulting from the nuclear data based on the uncertainty of the radiation source intensity based on the difference between an actual measured value of the amount of generation of nuclides and a calculated value from the calculation code and the uncertainty of the radiation source intensity derived from the nuclear data, thereby evaluating the uncertainty of the radiation source intensity calculated by the calculation code; evaluating the uncertainty of the radiation source intensity based on a condition for a fuel of a reactor core on the basis of a reactor core design database in which a parameter of the fuel of the reactor core is recorded; and calculating a radiation source intensity after correction obtained by correcting a radiation source intensity to be a predetermined reference with the uncertainty of the radiation source intensity calculated by the calculation code and the uncertainty of radiation source intensity based on the fuel condition.SELECTED DRAWING: Figure 1

Description

[Technical Field]

[0001] The present disclosure relates to a radiation source intensity evaluation method, a shielding analysis method, and a radiation source intensity evaluation device and program. [Background technology]

[0002] When designing shielding for facilities and casks that store spent fuel from nuclear power plants, the radiation source strength of the spent fuel is required. The radiation source strength is evaluated by calculation, but since there is no actual spent fuel to store at the time of shielding design, it is not possible to evaluate the exact value of the radiation source strength. Furthermore, it is not realistic to calculate the radiation source strength for each individual spent fuel. In order to perform accurate shielding design, it is necessary to evaluate the radiation source strength by taking into account uncertainties due to various factors, such as variations in the spent fuel. As a related technique, Patent Document 1 discloses a method for evaluating the decay heat of radioactive materials that constitute nuclear fuel, including the uncertainty involved. [Prior art documents] [Patent documents]

[0003] [Patent Document 1] Japanese Patent Application Publication No. 2020-76669 Summary of the Invention [Problem to be solved by the invention]

[0004] A method for evaluating the radiation source strength of spent fuel that takes into account the uncertainty is required.

[0005] The present disclosure provides a radiation source strength evaluation method, a shielding analysis method, a radiation source strength evaluation device, and a program that can solve the above-mentioned problems. [Means for solving the problem]

[0006] The method for evaluating radiation source strength according to the present disclosure includes the steps of: evaluating uncertainty of radiation source strength due to nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay and production of the nuclide are varied; evaluating uncertainty of radiation source strength due to the calculation code by calibrating the uncertainty of radiation source strength due to the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength due to the nuclear data; evaluating uncertainty of radiation source strength due to fuel conditions, based on a core design database in which parameters of the fuel in the core are recorded; and calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of radiation source strength due to the calculation code and the uncertainty of radiation source strength due to the fuel conditions. Furthermore, a method for evaluating radiation source strength according to the present disclosure includes the steps of: evaluating uncertainty of radiation source strength caused by nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay or production of the nuclide are varied; evaluating uncertainty of radiation source strength caused by the calculation code by calibrating the uncertainty of radiation source strength caused by the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual measured value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength caused by the nuclear data; and calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of the radiation source strength calculated by the calculation code. In the step of evaluating the uncertainty of the radiation source strength based on the calculation code, the uncertainty of the radiation source strength of the first nuclide is estimated based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, either by producing the first nuclide or by decay, etc. of the first nuclide, using a calculation formula that shows the relationship between the sensitivity that represents fluctuations of the first nuclide with respect to fluctuations in the number density of the second nuclide, the difference between the radiation source strength based on the actual value of the production amount of the second nuclide and the radiation source strength based on the calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide, and based on the relational formula between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, coefficients of the relational formula are calculated, and the uncertainty of the radiation source strength of the first nuclide is calibrated using the relational formula.

[0007] In the shielding analysis method according to the present disclosure, shielding analysis is performed using the corrected radiation source intensity calculated by the above evaluation method as an input.

[0008] The radiation source strength evaluation device according to the present disclosure comprises: means for evaluating uncertainty of radiation source strength caused by nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay and production of the nuclide are varied; means for evaluating uncertainty of radiation source strength caused by the calculation code by calibrating the uncertainty of radiation source strength caused by the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value of the calculation code, and the uncertainty of radiation source strength caused by the nuclear data; means for evaluating uncertainty of radiation source strength based on fuel conditions, based on a core design database in which parameters of the fuel in the core are recorded; and means for calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of radiation source strength caused by the calculation code and the uncertainty of radiation source strength based on the fuel conditions. The means for evaluating the uncertainty of the radiation source strength using the calculation code estimates the uncertainty of the radiation source strength of the first nuclide based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, either by producing the first nuclide or by decay, etc. of the first nuclide, using a calculation formula showing the relationship between the sensitivity that represents fluctuations of the first nuclide with respect to fluctuations in the number density of the second nuclide, the difference between the radiation source strength based on the actual measured value of the production amount of the second nuclide and the radiation source strength based on the calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide, calculates coefficients for the relational formula based on the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, and calibrates the uncertainty of the radiation source strength of the first nuclide using the relational formula.

[0009] The program according to the present disclosure includes the steps of: evaluating uncertainty of radiation source strength caused by nuclear data based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay and production of nuclides are varied; evaluating uncertainty of radiation source strength caused by the calculation code by calibrating the uncertainty of radiation source strength caused by the nuclear data based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength caused by the nuclear data; evaluating uncertainty of radiation source strength based on fuel conditions based on a core design database in which parameters of the fuel for the core are recorded; and calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of radiation source strength caused by the calculation code and the uncertainty of radiation source strength based on the fuel conditions. and in the step of evaluating the uncertainty of the radiation source strength based on the calculation code, the uncertainty of the radiation source strength of the first nuclide is estimated based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, the first nuclide being produced or which is produced by the decay or the like of the first nuclide, using a calculation formula showing the relationship between sensitivity representing fluctuations in the first nuclide with respect to fluctuations in the number density of the second nuclide, a difference between a radiation source strength based on an actual measurement value of the production amount of the second nuclide and a radiation source strength based on a calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide, and the uncertainty of the radiation source strength of the first nuclide, and the uncertainty of the radiation source strength of the first nuclide is calculated based on the relational formula between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, and the process of calibrating the uncertainty of the radiation source strength of the first nuclide using the relational formula. Execute the following. [Effects of the Invention]

[0010] According to the radiation source strength evaluation method, radiation source strength evaluation device, and program disclosed herein, it is possible to evaluate the radiation source strength including uncertainty. According to the shielding design method disclosed herein, it is possible to perform shielding design based on the radiation source strength evaluated including uncertainty. [Brief explanation of the drawings]

[0011] [Figure 1] FIG. 1 is a block diagram illustrating an example of an evaluation device according to an embodiment. [Figure 2A] FIG. 1 is a first diagram illustrating a process for evaluating uncertainty in radiation source intensity due to nuclear data according to an embodiment. [Figure 2B] FIG. 2 is a second diagram illustrating the process of evaluating the uncertainty of the radiation source intensity caused by nuclear data according to the embodiment. [Figure 3] FIG. 1 illustrates an example of a combustion chain according to an embodiment. [Figure 4A] FIG. 2 is a diagram showing an example of parameters registered in a core design database according to the embodiment. [Figure 4B] FIG. 2 is a diagram illustrating fuel selection according to an embodiment. [Figure 5] FIG. 10 is a diagram illustrating a dose rate taking uncertainty into consideration according to an embodiment. [Figure 6] 10 is a flowchart illustrating an example of a process for evaluating radiation source intensity according to the embodiment. [Figure 7] FIG. 2 illustrates an example of a hardware configuration of an evaluation apparatus according to an embodiment. DETAILED DESCRIPTION OF THE INVENTION

[0012] <Embodiment> The radiation source intensity evaluation device of the present disclosure will be described below with reference to FIGS. (composition) FIG. 1 is a block diagram illustrating an example of an evaluation device according to an embodiment. The evaluation device 10 calculates the radiation source intensity and dose rate of spent fuel, taking uncertainty into consideration. When evaluating the radiation source intensity and dose rate of spent fuel, a calculation code is used to calculate the production amount of nuclides (e.g., nuclides that emit radiation to be evaluated, such as neutrons and gamma rays) generated from the spent fuel, and the radiation source intensity, etc. are calculated from the production amount. Various parameters are input into the calculation code, but, for example, nuclear data among the input parameters contains errors and variations. Therefore, depending on the input values, variations occur in the production amount of nuclides calculated by the calculation code and the radiation source intensity and dose rate calculated based on the production amount of nuclides. In addition, there is a difference between the production amount of nuclides calculated by the calculation code and the production amount of nuclides actually generated from spent fuel. The evaluation device 10 evaluates [1] the uncertainty of the radiation source intensity due to the variability of nuclear data, and [2] the uncertainty of the radiation source intensity based on the nuclide production amount calculated by the calculation code (including the variability of nuclear data) (the difference between the calculated value of the radiation source intensity by the calculation code and the variability of the actual radiation source intensity).

[0013] Furthermore, when evaluating radiation source strength, input parameters such as reactor power and burnup, in addition to the nuclear data described above, are input into the calculation code to calculate the amount of nuclide generated and evaluate the radiation source strength and dose rate. However, in actual spent fuel, differences in radiation source strength occur due to differences in fuel specifications, irradiation history, plant operating conditions, etc. Regarding differences and variations based on irradiation history, etc., the evaluation device 10 extracts input parameters that are highly sensitive to radiation source strength from a core design database in which parameters of licensed cores and replacement cores are registered, and evaluates the uncertainty of radiation source strength caused by variations in the extracted input parameters. In this way, [3] the uncertainty of radiation source strength due to differences in fuel conditions such as irradiation history is evaluated. In this way, the evaluation device 10 evaluates [1] the uncertainty of the radiation source strength due to the nuclear data, [2] the uncertainty of the radiation source strength due to the calculation code, and [3] the uncertainty of the radiation source strength due to the input parameters related to the fuel conditions such as the irradiation history (hereinafter, this may be referred to as "input-induced uncertainty"). Taking these uncertainties [1] to [3] into consideration, the evaluation device 10 calculates the radiation source strength and dose rate of the spent fuel.

[0014] As shown in the figure, the evaluation device 10 includes an input receiving unit 11, a control unit 12, and a storage unit 13. The input receiving unit 11 receives information, instructions, etc. input using input devices such as a keyboard, a mouse, a touch panel, buttons, etc. The input receiving unit 11 records the received information in the storage unit 13 and outputs it to the control unit 12.

[0015] The control unit 12 executes processes such as the process of evaluating the uncertainties in [1] to [3] above, and the process of calculating the maximum value of the radiation source intensity and the dose rate taking the uncertainties in [1] to [3] into consideration. The control unit 12 also displays the processing results on a display device or outputs them as electronic data. The control unit 12 includes a first evaluation unit 121, a second evaluation unit 122, a third evaluation unit 123, and a shielding analysis unit 124. The functions of these units will be described later. The storage unit 13 stores calculation codes, a core design database, data currently being processed, and the like.

[0016] [1] Evaluation of uncertainties caused by nuclear data The first evaluation unit 121 [1] evaluates the uncertainty of the radiation source intensity due to nuclear data. FIG. 2A shows an example of nuclear data. The width e1 in FIG. 2A represents the width of the error and variation of the nuclear data. The nuclear data and its error are registered in a nuclear data library, and these values ​​can be used. The first evaluation unit 121 randomly selects multiple nuclear data within the range of width e1. For example, the first evaluation unit 121 selects nuclear data a1 to d1. Nuclear data b1 is a value for the reference state. The first evaluation unit 121 inputs the selected nuclear data a1 to d1 into a calculation code. The calculation code is, for example, the SCALE (Standardized Computer Analyses for Licensing Evaluation) system. When input parameters such as fuel assembly specification data (e.g., parameters related to nuclear data and irradiation history) are input into the SCALE system and executed, the SCALE system performs resonance calculations, transport calculations, burnup calculations, etc. to calculate the production amount of nuclides that emit gamma rays and neutrons. By specifying the irradiation period in the SCALE system, it is possible to calculate the nuclide production rate over time from irradiation to cooling. The first evaluation unit 121 inputs nuclear data a1 and other input parameters into a calculation code (SCALE system) and executes the calculation code. The first evaluation unit 121 calculates the source intensity by multiplying the nuclide production rate output by the calculation code by the decay constant. The first evaluation unit 121 performs similar calculations for nuclear data b1 to d1 and evaluates the source intensity for the nuclear data a1 to d1. Figure 2B shows an example of the source intensity evaluated for the nuclear data a1 to d1. The vertical axis of the graph in Figure 2B represents the source intensity, and the horizontal axis represents time. Source intensities a2, b2, c2, and d2 represent the transition of the source intensity corresponding to the nuclear data a1, b1, c1, and d1, respectively. The first evaluation unit 121 evaluates, for example, the uncertainty of the source intensity due to the nuclear data at time T (the time when fuel cooling is completed) in this graph. For example, the first evaluation unit 121 calculates the standard deviation based on the radiation source intensity b2 at time T when nuclear data b1 in the reference state is used, and sets the calculated standard deviation σ as the uncertainty of the radiation source intensity due to the nuclear data [1]. For example, in this example, ((a2 - b2) 2 +(c2-b2) 2+(d2-b2) 2 [1] Calculate the uncertainty of the source strength due to nuclear data by taking the square root of [1 / 3].

[0017] [2] Evaluation of uncertainty using calculation codes The second evaluation unit 122 evaluates the uncertainty of the radiation source strength due to the calculation code [2]. For example, for a certain nuclide A, if the uncertainty of the radiation source strength due to the nuclear data [1] evaluated by the first evaluation unit 121 is σ A On the other hand, the dispersion of the radiation source intensity (for example, the standard deviation of the difference between the measured value and the calculated value) based on the actual value of the production amount of nuclide A measured in the post-irradiation examination (PIE) conducted on the same fuel assembly as that used for analyzing the nuclide production amount by the calculation code is Δγ' A In this case, the second evaluation unit 122 determines α that satisfies the following equation (1): A and β A Estimate. Δγ´ A =α A σ A +β A ···(1) For example, Δγ´ A and σ A Prepare multiple combinations of α A and β A When evaluating the radiation source intensity of gamma rays, the second evaluation unit 122 may estimate α for the nuclide A1 among the multiple nuclides A1 to An that emit gamma rays. A1 and β A1 and for nuclide A2, α A2 and β A2 is estimated, and the process of ... is performed for each of the nuclides A1 to An. Then, the radiation source intensity of the gamma ray to be evaluated can be calculated by the following formula (2).

[0018]

number

[0019] The radiation source strength calculated by equation (2) is the radiation source strength that takes into account the uncertainty of the radiation source strength calculated by the calculation code [2]. calc is the source strength calculated by the conventional method. Δγ is the uncertainty of the source strength due to the calculation code [2]. As can be seen from the calculation process of Δγ above, Δγ includes the uncertainty of the source strength due to the nuclear code [1]. In the above example, the sigma term in equation (2) is the sum of the values ​​obtained by calibrating the source strength variations due to the nuclear data for nuclides A1 to An using equation (1). By adding the correction term Δγ, which is based on the uncertainty caused by errors in the nuclear data and the uncertainty caused by the difference between the calculation code and the actual measured value, the uncertainty of γ calculated by the conventional method can be reduced. calc It is possible to calculate a radiation source strength that is evaluated on the safe side compared to the above and has a reasonable basis. A1 , β A1 Once these are calculated, the uncertainty of the radiation source strength can be calculated using equation (2) [2].

[0020] (When actual measurements cannot be obtained in post-irradiation testing) However, there are cases where the actual measured value of the nuclide production amount to be evaluated cannot be measured in post-irradiation examination. In such cases, the burn-up chain is used to evaluate the radiation source strength of the unmeasured nuclide. Figure 3 shows an example of a burn-up chain. Nuclide A is the nuclide to be evaluated (e.g., a nuclide that serves as a gamma-ray source) that cannot be measured in post-irradiation examination. Nuclide B is an actinide nuclide that produces nuclide A by fission, and its production amount can be measured in post-irradiation examination. Nuclide C is a nuclide that is produced by neutron capture, decay, etc. of nuclide A, and its production amount can be measured in post-irradiation examination. Here, for nuclide i, which has a relationship of production, decay, etc. with nuclide A that can be described using the burn-up chain, the difference in radiation source strength based on the actual measured value of the nuclide production amount measured in post-irradiation examination and the radiation source strength based on the calculated value of the nuclide production amount calculated using a computer code is defined as Δγi. The fluctuation (%) in the number density of nuclide A when the number density of nuclide i, for which the actual measured value of the production amount is obtained, fluctuates by 1% is defined as W. i→A Let's say. W i→Acan be calculated using a calculation code. The dispersion and uncertainty of the source intensity of nuclide A is expressed as Δγ´ A Then, the following equation (3) can be derived from the combustion chain relationship shown in Figure 3, where nuclide A is produced from nuclide B, and nuclide C is produced from nuclide A.

[0021]

number

[0022] In the case of the burnup chain of FIG. 3, the second evaluation unit 122 calculates the uncertainty Δγ' of the burnup intensity of nuclide A, the amount of which cannot be measured, by setting i=B, C in the above equation (3). A Estimate.

[0023] In addition, in equation (3), i = A, B, C. This is because even if the amount of nuclide A produced can be measured, if the reliability of the value is low, it is possible to set i = A, B, C in the above equation (3) and obtain Δγ' A This is because by calculating the uncertainty of the burnup intensity of nuclide A, we can estimate it more accurately.

[0024] The second evaluation unit 122 calculates Δγ′ using equation (3). A When estimating the coefficient α using the above formula (1), it can be calculated in the same way as when the amount of nuclide produced can be measured in post-irradiation examination. A , β A Then, using the above equation (2), the source strength is calculated taking into account the uncertainty of the source strength calculated by the calculation code [2]. In this way, even if the amount of nuclide produced to be evaluated cannot be measured, it is possible to calculate the source strength taking into account the uncertainty of the source strength calculated by the calculation code [2] by using the relational equation (3) based on the burnup chain.

[0025] [3] Evaluation of input-related uncertainty The third evaluation unit 123 evaluates the uncertainty of the radiation source strength due to [3] input. We have previously explained methods for evaluating uncertainty based on the difference between the variance in nuclear data and the variance in the calculated and measured radiation source strength values. In contrast, the third evaluation unit 123 evaluates the uncertainty of the radiation source strength due to differences in fuel specifications, irradiation history, operating environment, etc. The core design database contains information about various cores, including licensed cores and replacement cores, such as reactor power, boron concentration, cooling period, the presence or absence of burnable poison (BP), and fuel composition. The third evaluation unit 123 extracts and analyzes parameters that affect the radiation source strength from the information about the fuel specifications, irradiation history, operating environment, etc. registered in the core design database, and evaluates the variance of the parameters that affect the radiation source strength. The third evaluation unit 123 classifies the parameters registered in the core design database into "H," "M," and "L" according to their degree of influence on the radiation source strength. "H" denotes a parameter with a high influence on the radiation source intensity, "M" denotes a parameter with a medium influence on the radiation source intensity, and "L" denotes a parameter with a low influence on the radiation source intensity. FIG. 4A shows an example of parameters classified as "H," "M," or "L" for gamma rays. Parameters classified as "H" include, for example, "power," parameters classified as "M" include, for example, "burnup," "composition," "presence or absence of BP," and "irradiation period," and parameters classified as "L" include, for example, "boron concentration," "adjacent fuel," "fuel temperature," and "moderator temperature." The "H," "M," and "L" classifications are set by knowledgeable engineers, and the third evaluation unit 123 extracts parameters from the core design database according to their influence on the radiation source intensity based on this setting.

[0026] After extracting parameters according to their influence on the radiation source intensity from the core design database, the third evaluation unit 123 sets a value for each parameter. For example, the third evaluation unit 123 sets a value randomly extracted from the core design database for parameters classified as "H" or "M." The third evaluation unit 123 may extract values ​​from the core design database so as to include, for example, the maximum and minimum values ​​of each parameter. Furthermore, the third evaluation unit 123 sets a representative value for parameters classified as "L." For example, the representative value for "fuel temperature" is set to 900 K, and the representative value for "moderator temperature" is set to a value described in the core characteristics manual. These values ​​are set by an engineer with knowledge, and the third evaluation unit 123 sets a representative value for each parameter according to this setting.

[0027] The third evaluation unit 123 sets randomly selected values ​​for the parameters "H" and "M" and a representative value for the parameter "L." The third evaluation unit 123 inputs these parameters into a calculation code to calculate, for example, the production yield of a nuclide emitting gamma rays. The third evaluation unit 123 then calculates the radiation source intensity for each nuclide by multiplying each calculation quantity by a decay constant. The third evaluation unit 123 repeatedly calculates the radiation source intensity by changing the values ​​set for the parameters "H" and "M" (where "L" is a representative value, i.e., a fixed value). This allows multiple radiation source intensities to be calculated. The third evaluation unit 123 evaluates the radiation source intensity variation based on the fuel specifications and irradiation history variations (i.e., [3] input-induced uncertainty) from the multiple calculated radiation source intensities. To ensure safety, it is necessary to evaluate the degree of variation when the radiation source intensity is high. Therefore, the third evaluation unit 123, for example, extracts calculation results with high radiation source intensity as described below and evaluates the variation for the fuel related to the extracted calculation results. Since the radiation source intensity is converted into a dose rate by shielding analysis, the radiation source intensity is converted into a dose rate below to evaluate the variation.

[0028] For example, to evaluate the uncertainty relative to the maximum dose rate of the licensed core, the third evaluation unit 123 selects a fuel with a high dose rate as the evaluation target. Figure 4B shows data obtained by converting the radiation source strength calculated by setting various values ​​for the "H" and "M" parameters into a dose rate. The horizontal axis of the graph in Figure 4B represents the difference between the calculated dose rate and the maximum dose rate of the licensed core, and the vertical axis represents the frequency of the calculation results corresponding to each difference. Data with a value of "0" on the horizontal axis represents the frequency of the calculation results when the dose rate is the same as the maximum dose rate of the licensed core. The third evaluation unit 123 selects a fuel with a high dose rate using this "0" as the reference. Range P1 represents the selected fuel. The third evaluation unit 123 converts the dose rate of the selected fuel into a radiation source strength, and evaluates the reliability coefficient of the source strength uncertainty using the converted radiation source strength as a sample using the bootstrap method. For example, the third evaluation unit calculates a confidence interval with a confidence coefficient of 95% (95% confidence level), and the width of this interval is set as (3) the uncertainty of the radiation source strength due to the input. The third evaluation unit also evaluates the uncertainty of the dose rate from the sensitivity of the dose rate to changes in the radiation source strength, the variation in the radiation source strength, and the confidence coefficient.

[0029] The shielding analysis unit 124 calculates a corrected radiation source intensity by correcting a predetermined reference radiation source intensity (for example, the radiation source intensity of a representative core such as a licensed core) using the uncertainty in the radiation source intensity due to the calculation code [2] evaluated by the second evaluation unit 122 and the uncertainty in the radiation source intensity due to the input [3] evaluated by the third evaluation unit 123. For example, the shielding analysis unit 124 calculates the corrected radiation source intensity by adding the maximum value of the uncertainty in the radiation source intensity due to the calculation code [2] and the maximum value of the uncertainty in the radiation source intensity due to the input [3] to the reference radiation source intensity. Then, the shielding analysis unit 124 uses the corrected radiation source intensity as an input to perform a shielding analysis of the spent fuel storage facility or cask to design the thickness of the shielding members, etc., and evaluate the dose rate.

[0030] Figure 5 shows the results of the dose rate evaluation. The horizontal axis of Figure 5 represents the dose rate. r0 is the maximum dose rate for the licensed core without considering uncertainty. r1 is the dose rate obtained by performing a shielding analysis using the source strength obtained by adding the uncertainty due to the [2] calculation code (including the uncertainty due to the nuclear code [1]) as input. r1 is the maximum dose rate for the licensed core based on reasonable grounds. r2 is the dose rate obtained by performing a shielding analysis using the source strength obtained by adding the uncertainty due to the [2] calculation code and the uncertainty due to the [3] input to the source strength corresponding to the maximum dose rate for the licensed core. r2 is the dose rate evaluation value taking uncertainty into account. Actual spent fuel differs from the licensed core in terms of specifications and irradiation history, but these differences are absorbed by the uncertainty due to the [3] input. Since performing shielding analysis of casks and the like using r0 is likely to affect safety, shielding analysis has conventionally been performed by estimating the radiation source strength overly conservatively. However, according to this embodiment, the uncertainty of the radiation source strength is reasonably evaluated by taking into account the uncertainty of the radiation source strength due to nuclear data and calculation codes, and the uncertainty of the radiation source strength due to variations in irradiation history, etc. This makes it possible to calculate a radiation source strength that is not overly conservative and is appropriately evaluated on the safe side. In fact, when comparing dose rates using conventional radiation source strengths with those using conventional radiation source strengths, it was confirmed that the dose rate evaluated using the radiation source strength evaluated by the method of this embodiment as input was lower than the dose rate evaluated using conventional conservative radiation source strengths as input.

[0031] (operation) Next, the flow of the evaluation process of the radiation source intensity in this embodiment will be described with reference to FIG. FIG. 6 is a flowchart showing an example of a process for evaluating radiation source intensity according to the embodiment. As a premise, the memory unit 13 stores a calculation code, a nuclear data library, a core design database, settings of "H", "M", and "L" for the input parameters to be input into the calculation code, and a representative value to be set for the "L" parameter. First, the operator specifies a nuclide that emits radiation (gamma rays, neutron rays, etc.) to be evaluated and instructs the evaluation of the radiation source strength. The input receiving unit 11 receives this instruction. The control unit 12 executes a process to evaluate the radiation source strength. First, the first evaluation unit 121 [1] evaluates the uncertainty of the radiation source strength due to nuclear data (step S1). For example, the first evaluation unit 121 randomly selects multiple values ​​for the reaction cross section of the nuclide to be evaluated from the error data in the nuclear data library, and executes a calculation code for the selected values ​​as input parameters to calculate the nuclide production amount. The first evaluation unit 121 converts the nuclide production amount obtained for the selected values ​​into radiation source strength and calculates the standard deviation due to the error data in the data library (uncertainty of the radiation source strength due to nuclear data). The first evaluation unit 121 calculates the uncertainty due to nuclear data for each nuclide and records it in the storage unit 13.

[0032] Next, the second evaluation unit 122 evaluates the uncertainty of the radiation source intensity by the calculation code [2] (step S2). When the amount of the nuclide produced to be evaluated can be measured in the post-irradiation examination, the second evaluation unit 122 calculates the dispersion (Δγ') of the radiation source intensity based on the actual measured value of the amount of the nuclide produced. A ) and the uncertainty of the source strength due to nuclear data (σ A ) based on the relation (the above equation (1)), the coefficient α A , β A The second evaluation unit 122 calculates the coefficient α A , β A Calculate Then, using equation (2), the uncertainty of the radiation source strength calculated by the calculation code [2] is calculated.

[0033] When the amount of nuclide produced as an evaluation target cannot be measured in the post-irradiation examination, the second evaluation unit 122 calculates the sensitivity (W ) that represents the fluctuation of nuclide A relative to the fluctuation of the number density of nuclide i (i=B, C) based on the relationship between nuclide A whose amount cannot be measured and measurable nuclide B that produces nuclide A or measurable nuclide C that is produced by the decay of nuclide A. i→A ) and the difference between the source intensity based on the measured value of the amount of nuclide i produced and the calculated value (Δγ i) and the uncertainty of the source strength of nuclide A (Δγ´ A ) (the above formula (3)) that indicates the relationship between the coefficient α A , β A and calculates the uncertainty of the radiation source strength by the calculation code [2] using equation (2). The second evaluation unit 122 records the calculated uncertainty of the radiation source strength by the calculation code in the storage unit 13.

[0034] Next, the third evaluation unit 123 evaluates the uncertainty of the radiation source intensity due to the input [3] (step S3). First, the third evaluation unit 123 classifies the parameters registered in the core design database into "H," "M," and "L." For parameters classified as "H" and "M," the third evaluation unit 123 sets values ​​randomly extracted from the core design database, and for parameters classified as "L," sets predetermined representative values. The third evaluation unit 123 combines the values ​​set for each parameter, inputs them into a calculation code, calculates the nuclide production amount, and calculates the radiation source intensity from the nuclide production amount. The third evaluation unit 123 calculates the nuclide production amount and the radiation source intensity multiple times by changing the combination of values ​​set for each parameter. This allows the radiation source intensity to be calculated according to the fuel specifications and variations in irradiation history. Next, the third evaluation unit 123 converts the calculated radiation source intensities into dose rates and selects a fuel indicated by a combination of parameter values ​​that results in a dose rate equal to or greater than a predetermined value, based on the maximum dose rate of a representative core, such as a licensed core. In the example of FIG. 4B, the fuel is selected such that the difference between the dose rate of the target fuel and the dose rate for licensed core burnup is equal to or greater than -0.05 μSv / h. (This value is an example and is not limiting.) The third evaluation unit 123 then replaces the dose rate of the selected fuel with the radiation source intensity and calculates the uncertainty of the radiation source intensity due to input (e.g., 95% confidence interval) using the selected fuel's radiation source intensity as a sample, using a bootstrap method or the like. The third evaluation unit 123 records the calculated uncertainty of the radiation source intensity due to input in the storage unit 13.

[0035] Next, the shielding analysis unit 124 evaluates the dose rate taking uncertainty into consideration (step S4). The shielding analysis unit 124 performs shielding analysis of the facilities and casks that store spent fuel by inputting the maximum value of the radiation source strength of the licensed core plus the maximum value of the uncertainty of the radiation source strength according to the calculation code recorded in the storage unit 13 and the maximum value of the uncertainty of the radiation source strength due to the input, and calculates a dose rate (r2 in FIG. 5) taking uncertainty into consideration based on the maximum dose rate of the licensed core (r0 in FIG. 5).

[0036] (effect) As described above, according to this embodiment, it is possible to calculate the radiation source strength of gamma rays and neutron rays while rationally considering uncertainties caused by the calculation code and inputs. This allows shielding design for spent fuel pools (SFPs) and casks to be performed based on radiation source strengths that are rationally and conservatively evaluated.

[0037] In the above embodiment, the radiation source strength is evaluated by adding up the uncertainties of [1] to [3] to the radiation source strength of the licensed core, but depending on the purpose of evaluating the radiation source strength, for example, the radiation source strength may be evaluated by adding up only the uncertainties of [1] and [2], or by adding up only the uncertainty of [3]. Also, the radiation source strength may be evaluated by adding up the uncertainties of [1] and [2] and uncertainties evaluated by a method other than [3], or by adding up the uncertainty of [3] and uncertainties evaluated by a method other than [1] and [2].

[0038] FIG. 7 is a diagram illustrating an example of a hardware configuration of the evaluation device. The computer 900 includes a CPU 901 , a main memory device 902 , an auxiliary memory device 903 , an input / output interface 904 , and a communication interface 905 . The evaluation device 10 described above is implemented in a computer 900. Each of the above-described functions is stored in the form of a program in an auxiliary storage device 903. A CPU 901 reads the program from the auxiliary storage device 903, loads it into a main storage device 902, and executes the above-described processing in accordance with the program. The CPU 901 also allocates a storage area in the main storage device 902 in accordance with the program. The CPU 901 also allocates a storage area in the auxiliary storage device 903 for storing data being processed in accordance with the program.

[0039] Alternatively, a program for implementing all or part of the functions of the evaluation device 10 may be recorded on a computer-readable recording medium, and the program may be loaded into a computer system and executed to perform processing by each functional unit. The term "computer system" as used herein includes hardware such as an OS and peripheral devices. Furthermore, if a WWW system is used, the term "computer system" also includes a homepage provision environment (or display environment). Furthermore, the term "computer-readable recording medium" refers to portable media such as CDs, DVDs, and USBs, as well as storage devices such as hard disks built into the computer system. Furthermore, if the program is distributed to the computer 900 via a communication line, the computer 900 that receives the program may load the program into the main storage device 902 and execute the above-described processing. Furthermore, the program may be for implementing part of the above-described functions, or may be capable of implementing the above-described functions in combination with a program already stored in the computer system.

[0040] As described above, several 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, substitutions, and modifications can be made without departing from the spirit of the invention. These embodiments and their modifications are included in the scope of the invention and its equivalents as defined in the claims, as well as in the scope and spirit of the invention.

[0041] <Additional Notes> The radiation source intensity evaluation method, shielding analysis method, radiation source intensity evaluation device, and program described in the embodiments can be understood, for example, as follows.

[0042] (1) A method for evaluating radiation source strength according to a first aspect includes the steps of: evaluating uncertainty of radiation source strength due to nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay or production of the nuclide are varied; evaluating uncertainty of radiation source strength due to the calculation code by calibrating the uncertainty of radiation source strength due to the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength due to the nuclear data; evaluating uncertainty of radiation source strength due to fuel conditions, based on a core design database in which parameters of the fuel in the core are recorded; and calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of radiation source strength due to the calculation code and the uncertainty of radiation source strength due to the fuel conditions. By calculating the radiation source strength taking into account uncertainties caused by the calculation code and input, it is possible to calculate the radiation source strength of gamma rays, neutrons, etc., taking into account the uncertainties used in shielding design in a reasonable manner.

[0043] (2) A method for evaluating radiation source strength according to a second aspect is the evaluation method of (1), wherein in the step of evaluating the uncertainty of radiation source strength using the calculation code, a coefficient of the relational equation (Equation (1)) between the variation of radiation source strength based on the actual measurement values ​​of the amount of nuclide produced and the uncertainty of radiation source strength caused by the nuclear data is calculated based on the relational equation, and the uncertainty is calibrated using the relational equation. This allows the uncertainty of the source strength to be calculated using the calculation code.

[0044] (3) A method for evaluating radiation source strength according to a third aspect is the evaluation method of (1), wherein in the step of evaluating the uncertainty of radiation source strength based on the calculation code, based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, either by producing the first nuclide or by decay, etc. of the first nuclide, the uncertainty of the radiation source strength of the first nuclide is estimated using a calculation formula (formula (3)) that shows the relationship between the sensitivity that represents fluctuations in the first nuclide relative to fluctuations in the number density of the second nuclide, the difference between the radiation source strength based on the actual value of the production amount of the second nuclide and the radiation source strength based on the calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide; based on the relational formula (formula (1)) between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, a coefficient of the relational formula is calculated, and the uncertainty of the radiation source strength of the first nuclide is calibrated using the relational formula. This makes it possible to calculate the uncertainty of the radiation source intensity using a calculation code for the nuclide, even if the nuclide that is the source of the radiation source intensity cannot be measured.

[0045] (4) A fourth aspect of the method for evaluating radiation source strength is the method of any one of (1) to (3), wherein in the step of evaluating the uncertainty of the radiation source strength based on the fuel conditions, parameters that affect the radiation source strength are varied and input to the calculation code, and the radiation source strength is calculated from the amount of nuclide produced output by the calculation code, thereby evaluating the uncertainty of the radiation source strength. This allows us to calculate the uncertainty of the source intensity based on various fuel specifications and irradiation histories.

[0046] (5) A fifth aspect of the method for evaluating radiation source strength is the method of any one of (1) to (4), wherein the step of evaluating the uncertainty of the radiation source strength based on the conditions of the fuel evaluates the uncertainty of the radiation source strength based on the parameters related to fuel whose radiation source strength is equal to or greater than a predetermined value. This allows the radiation source strength to be evaluated on the safe side.

[0047] (6) A sixth aspect of the method for evaluating radiation source strength is the method of evaluation according to any one of (1) to (5), in which the correction is performed by adding the maximum value of the reference radiation source strength to the maximum value of the uncertainty of the radiation source strength based on the calculation code and the maximum value of the uncertainty of the radiation source strength based on the fuel conditions. This allows the radiation source strength to be calculated while taking into account the uncertainty in a reasonable and safe manner.

[0048] (7) A seventh aspect of the method for evaluating radiation source strength is the method according to any one of (1) to (6), wherein the nuclide is a source of gamma rays or neutron rays. This makes it possible to evaluate the source strength of gamma rays, which are the subject of evaluation in the shielding analysis of the SFP, and gamma rays and neutron rays, which are the subject of evaluation in the shielding analysis of the cask.

[0049] (8) A method for evaluating radiation source strength according to an eighth aspect includes the steps of: evaluating uncertainty of radiation source strength caused by nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay or production of the nuclide are varied; evaluating uncertainty of radiation source strength caused by the calculation code by calibrating the uncertainty of radiation source strength caused by the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength caused by the nuclear data; and calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of radiation source strength caused by the calculation code. This allows us to provide source strength that takes into account the uncertainty of the calculation code.

[0050] (9) A method for evaluating radiation source strength according to a ninth aspect includes the steps of: evaluating uncertainty of radiation source strength based on the irradiation history of fuel based on a core design database in which parameters of the fuel of the core are recorded; and calculating a corrected radiation source strength by correcting a predetermined reference radiation source strength by the uncertainty of the radiation source strength based on the irradiation history. This makes it possible to provide source intensity that takes into account input-related uncertainties.

[0051] (10) A shielding analysis method according to a tenth aspect performs shielding analysis using the corrected radiation source intensity calculated by the radiation source intensity evaluation method according to any one of (1) to (9) as an input. By performing shielding analysis using radiation source strength as input, taking into account uncertainties caused by the calculation code and input, it is possible to accurately design shielding for spent fuel storage facilities and storage containers based on radiation source strength evaluated on the safe side.

[0052] (11) An evaluation device according to an eleventh aspect comprises: means for evaluating uncertainty of radiation source strength caused by nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay or production of the nuclide are varied; means for evaluating uncertainty of radiation source strength caused by the calculation code by calibrating the uncertainty of radiation source strength caused by the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength caused by the nuclear data; means for evaluating uncertainty of radiation source strength caused by fuel conditions, based on a core design database in which parameters of the fuel for the core are recorded; and means for calculating a corrected radiation source strength, by correcting a predetermined reference radiation source strength by the uncertainty of radiation source strength caused by the calculation code and the uncertainty of radiation source strength caused by the fuel conditions.

[0053] (12) A program according to a twelfth aspect causes a computer to execute the following steps: evaluating uncertainty of radiation source strength caused by nuclear data, based on the amount of nuclide produced calculated by a calculation code when values ​​of nuclear data input into the calculation code for calculating the decay or production of the nuclide are varied; evaluating uncertainty of radiation source strength caused by the calculation code by calibrating the uncertainty of radiation source strength caused by the nuclear data, based on the uncertainty of radiation source strength based on the difference between the actual value of the amount of nuclide produced and the calculated value by the calculation code, and the uncertainty of radiation source strength caused by the nuclear data; evaluating uncertainty of radiation source strength based on fuel conditions, based on a core design database in which parameters of the fuel for the core are recorded; and calculating a corrected radiation source strength, in which a predetermined reference radiation source strength is corrected by the uncertainty of radiation source strength caused by the calculation code and the uncertainty of radiation source strength based on the fuel conditions. [Explanation of symbols]

[0054] 10. Evaluation device 11 Input reception section 12 Control section 121···First evaluation section 122...Second evaluation section 123···Third Evaluation Section 124...Occupation analysis section 13...Storage section 900···Computer 901 CPU 902...Main memory 903...Auxiliary storage device 904 Input / Output Interface 905···Communication Interface

Claims

1. a step of evaluating the uncertainty of the radiation source intensity caused by the nuclear data based on the amount of nuclide produced calculated by a calculation code for calculating the decay and production of the nuclide when values ​​of the nuclear data input into the calculation code are varied; a step of evaluating the uncertainty of the radiation source strength calculated by the calculation code by calibrating the uncertainty of the radiation source strength caused by the nuclear data based on the uncertainty of the radiation source strength based on the difference between the actual measurement value of the amount of nuclide produced and the calculated value of the calculation code, and the uncertainty of the radiation source strength caused by the nuclear data; A step of evaluating the uncertainty of the source strength based on the conditions of the fuel based on a core design database in which parameters of the fuel of the core are recorded; a step of calculating a corrected radiation source intensity by correcting a predetermined reference radiation source intensity by the uncertainty of the radiation source intensity determined by the calculation code and the uncertainty of the radiation source intensity based on the fuel conditions; and In the step of evaluating the uncertainty of the radiation source strength based on the calculation code, based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, either producing the first nuclide or resulting from the decay or the like of the first nuclide, the uncertainty of the radiation source strength of the first nuclide is estimated using a calculation formula showing the relationship between sensitivity representing fluctuations of the first nuclide with respect to fluctuations in number density of the second nuclide, a difference between a radiation source strength based on an actual measured value of the production amount of the second nuclide and a radiation source strength based on a calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide; calculating a coefficient of a relational expression based on the relational expression between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, and calibrating the uncertainty of the radiation source strength of the first nuclide using the relational expression; Methods for evaluating radiation source strength.

2. In the step of evaluating the uncertainty of the radiation source intensity based on the fuel conditions, parameters that affect the radiation source intensity are varied and input into the calculation code, and the radiation source intensity is calculated from the amount of nuclide produced output by the calculation code, thereby evaluating the uncertainty of the radiation source intensity. The method for evaluating radiation source intensity according to claim 1 .

3. In the step of evaluating the uncertainty of the radiation source intensity based on the fuel conditions, the uncertainty of the radiation source intensity is evaluated based on the parameters related to fuel whose radiation source intensity is equal to or greater than a predetermined value. The method for evaluating radiation source intensity according to claim 2.

4. In the correction, a maximum value of uncertainty of the radiation source intensity based on the calculation code and a maximum value of uncertainty of the radiation source intensity based on the fuel conditions are added to a maximum value of the reference radiation source intensity to perform the correction. The method for evaluating radiation source intensity according to any one of claims 1 to 3.

5. The nuclide is a source of gamma rays or neutron rays; The method for evaluating radiation source intensity according to any one of claims 1 to 3.

6. a step of evaluating the uncertainty of the radiation source intensity caused by the nuclear data based on the amount of nuclide produced calculated by a calculation code for calculating the decay and production of the nuclide when values ​​of the nuclear data input into the calculation code are varied; a step of evaluating the uncertainty of the radiation source strength calculated by the calculation code by calibrating the uncertainty of the radiation source strength caused by the nuclear data based on the uncertainty of the radiation source strength based on the difference between the actual measurement value of the amount of nuclide produced and the calculated value of the calculation code, and the uncertainty of the radiation source strength caused by the nuclear data; a step of calculating a corrected radiation source intensity by correcting a predetermined reference radiation source intensity by the uncertainty of the radiation source intensity calculated by the calculation code; and In the step of evaluating the uncertainty of the radiation source strength based on the calculation code, based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, either producing the first nuclide or resulting from the decay or the like of the first nuclide, the uncertainty of the radiation source strength of the first nuclide is estimated using a calculation formula showing the relationship between sensitivity representing fluctuations of the first nuclide with respect to fluctuations in number density of the second nuclide, a difference between a radiation source strength based on an actual measured value of the production amount of the second nuclide and a radiation source strength based on a calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide; calculating a coefficient of a relational expression based on the relational expression between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, and calibrating the uncertainty of the radiation source strength of the first nuclide using the relational expression; Methods for evaluating radiation source strength.

7. performing a shielding analysis using the corrected radiation source intensity calculated by the evaluation method according to any one of claims 1 to 3 as an input; Occlusion analysis method.

8. a means for evaluating the uncertainty of the radiation source intensity caused by the nuclear data based on the amount of nuclide produced calculated by the calculation code when the value of the nuclear data input to the calculation code for calculating the decay and production of the nuclide is varied; a means for evaluating the uncertainty of the radiation source strength due to the calculation code by calibrating the uncertainty of the radiation source strength due to the nuclear data based on the uncertainty of the radiation source strength due to the difference between the actual measurement value of the amount of nuclide produced and the calculated value of the calculation code, and the uncertainty of the radiation source strength due to the nuclear data; a means for evaluating the uncertainty of the radiation source strength based on the conditions of the fuel based on a core design database in which parameters of the fuel in the core are recorded; a means for calculating a corrected radiation source intensity by correcting a predetermined reference radiation source intensity by the uncertainty of the radiation source intensity determined by the calculation code and the uncertainty of the radiation source intensity based on the fuel conditions; and The means for evaluating the uncertainty of the radiation source strength using the calculation code estimates the uncertainty of the radiation source strength of the first nuclide based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, which produces the first nuclide or is produced by the decay or the like of the first nuclide, using a calculation formula showing the relationship between the sensitivity representing fluctuations of the first nuclide with respect to fluctuations in number density of the second nuclide, the difference between the radiation source strength based on the actual measured value of the production amount of the second nuclide and the radiation source strength based on the calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide; calculating a coefficient of a relational expression based on the relational expression between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, and calibrating the uncertainty of the radiation source strength of the first nuclide using the relational expression; A radiation source strength evaluation device.

9. On the computer, a step of evaluating the uncertainty of the radiation source intensity caused by the nuclear data based on the amount of nuclide produced calculated by a calculation code for calculating the decay and production of the nuclide when values ​​of the nuclear data input into the calculation code are varied; a step of evaluating the uncertainty of the radiation source strength calculated by the calculation code by calibrating the uncertainty of the radiation source strength caused by the nuclear data based on the uncertainty of the radiation source strength based on the difference between the actual measurement value of the amount of nuclide produced and the calculated value of the calculation code, and the uncertainty of the radiation source strength caused by the nuclear data; A step of evaluating the uncertainty of the source strength based on the conditions of the fuel based on a core design database in which parameters of the fuel of the core are recorded; a step of calculating a corrected radiation source intensity by correcting a predetermined reference radiation source intensity by the uncertainty of the radiation source intensity determined by the calculation code and the uncertainty of the radiation source intensity based on the fuel conditions; and In the step of evaluating the uncertainty of the radiation source strength based on the calculation code, based on the relationship between a first nuclide whose production amount cannot be measured and a second nuclide whose production amount can be measured, either producing the first nuclide or resulting from the decay or the like of the first nuclide, the uncertainty of the radiation source strength of the first nuclide is estimated using a calculation formula showing the relationship between sensitivity representing fluctuations of the first nuclide with respect to fluctuations in number density of the second nuclide, a difference between a radiation source strength based on an actual measured value of the production amount of the second nuclide and a radiation source strength based on a calculated value of the production amount of the second nuclide, and the uncertainty of the radiation source strength of the first nuclide; a process of calculating a coefficient of a relational expression based on the relational expression between the estimated uncertainty of the radiation source strength of the first nuclide and the uncertainty of the radiation source strength caused by the nuclear data for the first nuclide, and calibrating the uncertainty of the radiation source strength of the first nuclide using the relational expression; A program that executes the following.

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