Analysis system, analysis method, and analysis program

Abstract

To provide an analysis system, an analysis method, and an analysis program capable of precisely estimating the minimum critical heat flux ratio.SOLUTION: An analysis system for performing an analysis of a safety evaluation when a main steam pipe provided in nuclear power facilities includes an operation part for estimating the minimum critical heat flux ratio as a parameter of a safety evaluation by using an evaluation model, the evaluation model is a model for expressing a correlation relationship among a critical heat flux ratio, an area in a predetermined section in an axial output distribution of a fuel rod of a fuel assembly loaded to a reactor core, and at least one of a cooling material temperature in a cooling material inlet side of a fuel assembly and a radial output peak value in a fuel assembly, and the operation part executes a step of acquiring at least one of an area, and a cooling material temperature, and a radial output peak value, and a step of calculating the minimum critical heat flux ratio when it is abnormal from an evaluation model on the basis of an area, a cooling material temperature, and a radial output peak value.SELECTED DRAWING: Figure 4

Description

[Technical Field]

[0001] The present disclosure relates to an analysis system, an analysis method, and an analysis program. [Background technology]

[0002] Conventionally, as a core design, a nuclear reactor core design support system is known that evaluates the reactor shutdown margin and thermal limit values ​​for a core design proposal (see, for example, Patent Document 1). [Prior art documents] [Patent documents]

[0003] [Patent Document 1] Japanese Patent Application Laid-Open No. 2007-147529 Summary of the Invention [Problem to be solved by the invention]

[0004] In the core design of a pressurized water reactor, one of the safety assessments is to ensure that the minimum critical heat flux ratio (DNBR) in the event of a main steam line rupture accident satisfies the limit value. DNBR (Departure from Nuclear Boiling Ratio) is the ratio of the critical heat flux to the actual local heat flux. In the analysis for the licensing application (license analysis), it is necessary to demonstrate that the minimum DNBR satisfies the limit value under conditions where various parameters are conservatively set for a representative core.

[0005] Here, in a core design such as that of Patent Document 1, for example, if an estimated value of the minimum DNBR is obtained as an evaluation result, the accuracy of the estimated value of the minimum DNBR is low, in other words, if the certainty of the estimated value of the minimum DNBR is low, it is not possible to determine logic for setting core conditions with strict minimum DNBR, and therefore explainability in licensing cannot be ensured.

[0006] Therefore, an object of the present disclosure is to provide an analysis system, an analysis method, and an analysis program that can accurately estimate the minimum critical heat flux ratio. [Means for solving the problem]

[0007] The analysis system disclosed herein is an analysis system that analyzes safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility, and includes a calculation unit that estimates a minimum critical heat flux ratio, which is a parameter for safety assessment, using an evaluation model. The evaluation model is a model that represents a correlation between the critical heat flux ratio, the area of ​​a predetermined section in the axial power distribution of fuel rods in a fuel assembly loaded in a reactor core, and at least one of the coolant temperature at the coolant inlet side of the fuel assembly and the radial power peak value of the fuel assembly. The calculation unit executes the steps of acquiring the area and at least one of the coolant temperature and the radial power peak value, and calculating the minimum critical heat flux ratio in the event of the abnormality from the evaluation model based on the area and at least one of the coolant temperature and the radial power peak value.

[0008] The analysis method disclosed herein is an analysis method executed by an analysis system that analyzes safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility, wherein the analysis system estimates a minimum critical heat flux ratio, which is a parameter for safety assessment, using an evaluation model, and the evaluation model is a model that represents a correlation between the critical heat flux ratio, the area of ​​a predetermined section in the axial power distribution of fuel rods in a fuel assembly loaded in a reactor core, and at least one of the coolant temperature at the coolant inlet side of the fuel assembly and the radial power peak value of the fuel assembly, and executes the steps of acquiring the area and at least one of the coolant temperature and the radial power peak value, and calculating the minimum critical heat flux ratio in the event of the abnormality from the evaluation model based on the area and at least one of the coolant temperature and the radial power peak value.

[0009] The analysis program disclosed herein is an analysis program executed by an analysis system that analyzes safety assessments in the event of an abnormality in a main steam pipe installed in a nuclear facility, and the analysis system estimates a minimum critical heat flux ratio, which is a parameter for safety assessment, using an evaluation model. The evaluation model is a model that represents a correlation between the critical heat flux ratio, the area of ​​a predetermined section in the axial power distribution of fuel rods in a fuel assembly loaded in a reactor core, and at least one of the coolant temperature at the coolant inlet side of the fuel assembly and the radial power peak value in the fuel assembly. The analysis program executes the following steps: acquiring the area and at least one of the coolant temperature and the radial power peak value; and calculating the minimum critical heat flux ratio in the event of the abnormality from the evaluation model based on the area and at least one of the coolant temperature and the radial power peak value. [Effects of the Invention]

[0010] According to the present disclosure, the minimum critical heat flux ratio can be estimated with high accuracy. [Brief explanation of the drawings]

[0011] [Figure 1] FIG. 1 is a schematic diagram of a nuclear reactor to which the present invention is applied. [Figure 2] FIG. 2 is a schematic diagram showing an example of the configuration of the analysis system according to this embodiment. [Figure 3] FIG. 3 is a diagram showing the axial power distribution of the fuel rods. [Figure 4] FIG. 4 is a flowchart showing an analysis method for calculating the minimum DNBR in an abnormal state. [Figure 5] FIG. 5 is a diagram comparing the estimated minimum DNBR values ​​calculated by the conventional analysis method and the estimated minimum DNBR values ​​calculated by the analysis method of this embodiment. DETAILED DESCRIPTION OF THE INVENTION

[0012] Hereinafter, embodiments of the present disclosure will be described in detail with reference to the drawings. However, this disclosure is not limited to these embodiments. Furthermore, the components in the following embodiments include those that are easily replaceable by those skilled in the art, or those that are substantially identical. Furthermore, the components described below can be combined as appropriate, and when there are multiple embodiments, the respective embodiments can also be combined.

[0013] [Present embodiment] (Reactor configuration) FIG. 1 is a schematic diagram of a nuclear reactor to which the present invention is applied.

[0014] Although not shown, the nuclear power plant has a nuclear reactor, a steam generator, and a steam turbine power generation facility located within a nuclear containment vessel. The nuclear reactor according to this embodiment uses light water as a reactor coolant and a neutron moderator, and maintains a high-temperature, high-pressure state that does not boil throughout the reactor core. This high-temperature, high-pressure light water is sent to the steam generator to generate steam through heat exchange, and this steam is then sent to a turbine generator to generate electricity. This is a pressurized water reactor (PWR).

[0015] A nuclear reactor heats the primary coolant through nuclear fission of fuel, and a steam generator exchanges heat between this high-temperature, high-pressure primary coolant and the secondary coolant to generate high-pressure steam.Steam turbine power generation equipment generates electricity by using this steam to drive a steam turbine.Meanwhile, the steam that drives the steam turbine is cooled in a condenser and condensed into water, which is returned to the steam generator.

[0016] As shown in FIG. 1, in a pressurized water reactor 10, a reactor vessel 11 is composed of a reactor vessel main body 12 and a reactor vessel lid 13 attached to the top of the reactor vessel main body 12, and the reactor vessel lid 13 is fixed to the reactor vessel main body 12 by a plurality of stud bolts and nuts so as to be able to be opened and closed.

[0017] The reactor vessel body 12 has a cylindrical shape with a closed vertical lower portion, and is provided with an inlet nozzle 14 for supplying light water as primary coolant at its upper vertical portion, and an outlet nozzle 15 for discharging the light water. The reactor vessel body 12 has a core barrel 16 disposed therein, and the upper portion of the core barrel 16 is supported by the inner wall surface of the reactor vessel body 12.

[0018] The core 17 is constructed by arranging a large number of fuel assemblies 20 as nuclear fuel in an area partitioned by an upper core 18 and a lower core 19 in the core barrel 16. The fuel assemblies 20 are constructed by bundling a plurality of fuel rods (not shown) along the vertical direction in a lattice pattern. The core 17 has a large number of control rods (not shown) arranged within the fuel assemblies 20, and these control rods are inserted into and removed from the core 17 by a control rod drive mechanism 21 to control the reactor output.

[0019] The core 17 to be analyzed in this embodiment is the core of a pressurized water reactor as described above, but the structure of the core 17 to be analyzed is not limited to the above description and may be any structure, such as the core of a boiling water reactor.

[0020] (Analysis system) 2 is a schematic diagram showing an example of the configuration of an analysis system according to this embodiment. The analysis system 30 of this embodiment uses an evaluation model to calculate an estimated value of the minimum DNBR (minimum critical heat flux ratio) in the event of an abnormality such as a main steam line break accident.

[0021] Here, DNBR refers to the ratio of the critical heat flux (CHF) at which heat transfer from the fuel cladding to the reactor coolant decreases and the cladding temperature begins to rise, to the actual local power. In the nucleate boiling region, the cladding temperature is sufficiently low, but when the CHF is reached, nucleate boiling cannot be maintained and transition occurs to film boiling, where the heated surface is covered with a vapor film. Since film boiling reduces heat transfer efficiency, the heat flux at the transition point is called the CHF and is managed accordingly. Furthermore, the minimum DNBR is the minimum value of the DNBR of the fuel rod that is most thermally severe in the reactor core. Next, the configuration of the analysis system 30 will be specifically described.

[0022] The analysis system 30 has a calculation unit 31, a storage unit 32, an input unit 33, and an output unit 34. The analysis system 30 may be configured as a single device, or may be configured as a plurality of devices combining a calculation device and a data server, etc., and is not particularly limited.

[0023] The calculation unit 31 includes an integrated circuit such as a CPU (Central Processing Unit), etc. The calculation unit 31 executes various programs and performs various processes for calculating an estimated value of the minimum DNBR.

[0024] The memory unit 32 is any memory device, such as a semiconductor memory device or a magnetic memory device. The memory unit 32 stores various programs for executing various processes and various data used in the processes. Examples of the various data include a fuel loading pattern related to a proposed replacement core design. The fuel loading pattern indicates the state of the fuel assemblies loaded, and includes information about the fuel assemblies of the replacement core, such as the number of new fuel assemblies, the fuel loading position, and the control rod pattern. The various data also include the results of a core analysis corresponding to the fuel loading pattern. The analysis results may be acquired from outside the analysis system 30 or analyzed by the analysis system 30. The results of the core analysis include the axial power distribution of the fuel rods, the coolant temperature at the coolant inlet side (inlet nozzle 14 side) of the fuel assembly 20, and the radial power peak value of the fuel assembly 20 when evaluating a condition in which all control rods except one control rod cluster are inserted and the reactor is in a hot shutdown state. The various programs include, for example, an analysis program P for calculating an estimated value of the minimum DNBR. The analysis program P includes the following evaluation model.

[0025] The input unit 33 is, for example, an input device such as a keyboard and a mouse. The output unit 34 is, for example, a display device such as a liquid crystal display. The input unit 33 and the output unit 34 may be integrated into an input / display device that allows input operations, such as a touch panel.

[0026] Incidentally, the core state when evaluating the condition in which all control rods except one control rod cluster are inserted and the reactor is in a thermal shutdown state is zero power and one control rod is stuck. On the other hand, the core state during an abnormality is assumed to be a low power and one control rod is stuck, and the axial power distribution is distorted toward the upper side of the core. Therefore, the core state when all control rods except one control rod cluster are inserted and the reactor is in a thermal shutdown state is similar to the core state during an abnormality. The area F of a specified section of the axial power distribution of fuel rods when all control rods except one control rod cluster are inserted and the reactor is in a thermal shutdown state is top Since there is a correlation between the minimum DNBR under abnormal conditions and the minimum DNBR under abnormal conditions, an evaluation model is created to calculate an estimated value of the minimum DNBR under abnormal conditions using this correlation. At this time, the evaluation model includes input parameters for more accurate evaluation of the reactor core state.

[0027] Here, the evaluation model is explained. The evaluation model is based on the DNBR and the area F top is a regression equation obtained from the correlation between the coolant temperature on the coolant inlet side and the radial power peak value in the fuel assembly 20. The regression equation is expressed by equation (1).

[0028] Estimated DNBR = (a·F top +b)(c·F ΔH +d)(e·T in / T in0 +f) ···(1) DNBR: Critical heat flux ratio F top :area F ΔH : Radial output peak value T in : Coolant temperature at the inlet after the abnormality T in0 : Average coolant temperature at the inlet before the abnormality a,b,c,d,e,f:Constant

[0029] As shown in equation (1), the evaluation model is based on the area F top and coolant temperature Tin , T in0 and the radial output peak value F ΔH The model uses a and b as input parameters and outputs an estimated value of DNBR. The evaluation model in equation (1) is generated by, for example, regression analysis such as the least squares method, while varying the constants a to f, so that it matches the analysis results. In other words, the evaluation model sets the constants so that the regression equation fits the analysis results. Alternatively, the regression equation may be obtained using principal component regression, Deming regression, geometric mean regression, or Passing-Pablok regression.

[0030] Area F top is the area of ​​a given section in the axial direction of the fuel rod, and is calculated based on the axial power distribution of the fuel rod when evaluating the condition in which all control rods except for one control rod cluster obtained as an analysis result are inserted and the reactor is in a hot shutdown state. Here, the power distribution refers to the spatial distribution of thermal power within the reactor core, and the axial power distribution refers to the distribution of thermal power in the longitudinal direction of the fuel rod.

[0031] A three-dimensional core analysis code is used to calculate the power distribution. Note that the above analysis code is not particularly limited, and any analysis code may be used.

[0032] FIG. 3 is a diagram showing the axial power distribution of the fuel rod. The horizontal axis indicates the power of the fuel rod, and the vertical axis indicates the axial height of the fuel rod. Here, the area of ​​the hatched portion over a predetermined axial section in the axial power distribution of the fuel rod shown in FIG. 3 is defined as F. top In other words, the area F top is the area of ​​the axial power distribution in the section from the upper end 100, which is the upper end in the vertical direction of the axial direction of the fuel rod, to the intermediate section 200, which is a predetermined distance away vertically downward.

[0033] Area F top may be calculated using the following formula (2):

[0034] (Area F top ) = (average power of a given section of fuel rods) / (average core power) (2)

[0035] The area F of a given section in the axial direction of the axial power distribution of the fuel rod top The predetermined section for determining the above may be a section extending from the upper end 100 in the vertical direction along the axial direction to the intermediate section 200, which is located vertically downward along the axial direction at a distance of ¼ to ½ of the total axial length of the fuel rod.

[0036] The coolant temperature at the coolant inlet side and the radial power peak value are the values ​​obtained as analysis results when all control rods except for one control rod cluster are inserted and the reactor is in a high-temperature shutdown state.

[0037] In such an analysis system 30, the calculation unit 31 calculates the area F top a parameter acquisition process for acquiring input parameters for calculating the minimum DNBR; and a minimum DNBR calculation process for calculating the minimum DNBR.

[0038] The area calculation process calculates the area F of a predetermined section in the axial direction of the fuel rod based on the axial power distribution of the fuel rod in the core state under the abnormal condition to be evaluated. top Calculate the area F top is the area F top The calculation method is the same as

[0039] The parameter acquisition process acquires input parameters for calculating the minimum DNBR. The input parameters are the area F calculated by the area calculation process. top and the coolant temperature on the coolant inlet side and the radial power peak value in the fuel assembly 20, which are obtained as analysis results. The coolant temperature and the radial power peak value are analysis results for the core state under abnormal conditions to be evaluated.

[0040] In the minimum DNBR calculation process, the input parameters acquired in the parameter acquisition process are input to the evaluation model of equation (1) to calculate the minimum DNBR in the event of an abnormality.

[0041] The calculation unit 31 performs a safety evaluation based on the calculated minimum DNBR. For example, the calculation unit 31 may determine whether the calculated minimum DNBR is equal to or greater than a predetermined limit value to determine whether the safety is acceptable.

[0042] Next, an analysis method for calculating the minimum DNBR by the analysis system 30 will be described with reference to Fig. 4. In this analysis method, the calculation unit 31 executes the analysis program P, thereby performing various steps to be described later. First, in the analysis method, the calculation unit 31 executes an area calculation process to obtain the axial power distribution of the fuel rods in the core state at the time of an abnormality based on the fuel loading pattern, and calculates the area F based on the obtained axial power distribution of the fuel rods. top is calculated (step S1).

[0043] Next, the calculation unit 31 executes a parameter acquisition process to obtain the area F calculated in step S1. top Then, the coolant temperature on the coolant inlet side and the radial power peak value in the fuel assembly 20 are acquired as the analysis results (step S2).

[0044] Thereafter, the calculation unit 31 executes a minimum DNBR calculation process, inputs the acquired input parameters into the evaluation model, calculates the minimum DNBR (step S3), and ends the process in the analysis method.

[0045] Next, the conventional evaluation results and the evaluation results of this embodiment will be compared with each other with reference to Fig. 5. In Fig. 5, the horizontal axis represents the evaluation index (i.e., the estimated value of DNBR), and the vertical axis represents the analysis result of DNBR. In the conventional evaluation index, the input parameter is the area F top In contrast to the evaluation model using only the above formula (1), the evaluation index of this embodiment is the evaluation model using the above formula (1).

[0046] In FIG. 5, the graph on the left shows the conventional evaluation index, and the graph on the right shows the evaluation index of this embodiment. Furthermore, the conventional and present evaluation indexes are obtained by acquiring multiple evaluation results based on conditions when the coolant temperature is varied. In the graph on the left of FIG. 5, the evaluation results, i.e., the estimated values ​​of DNBR, vary. In contrast, in the graph on the right of FIG. 5, the evaluation results, i.e., the estimated values ​​of DNBR, are suppressed. From the above, it was confirmed that the minimum DNBR can be accurately estimated by using the evaluation model of this embodiment.

[0047] In this embodiment, the evaluation model includes the coolant temperature and the radial power peak value as input parameters, but it is sufficient if it includes at least one of the input parameters.

[0048] As described above, the analysis system 30, the analysis method, and the analysis program P described in this embodiment can be understood, for example, as follows.

[0049] The analysis system 30 according to the first aspect is an analysis system 30 for analyzing safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility, and includes a calculation unit 31 for estimating a minimum critical heat flux ratio, which is a parameter of the safety assessment, using an evaluation model. The evaluation model is based on the critical heat flux ratio and the area F of a predetermined section in the axial power distribution of the fuel rods of the fuel assembly 20 loaded in the core. top and the coolant temperature T at the coolant inlet side of the fuel assembly 20. in , T in0 and the radial power peak value F in the fuel assembly 20 ΔH The calculation unit 31 calculates the area F top and the coolant temperature T in , T in0 and the radial output peak value F ΔH and a step S2 of acquiring at least one of the area F top and the coolant temperature T in , T in0 and the radial output peak value FΔH and Step S3 of calculating the minimum critical heat flux ratio at the time of the abnormality from the evaluation model based on at least one of the above.

[0050] According to this configuration, the area F top and coolant temperature T in , T in0 and radial power peak value F ΔH By inputting input parameters including at least one of the above into the evaluation model, it is possible to calculate an accurate estimate of the minimum DNBR.

[0051] As a second aspect, in the analysis system 30 according to the first aspect, the evaluation model is expressed by a regression equation, The regression equation is Estimated DNBR = (a·F top +b)(c·F ΔH +d)(e·T in / T in0 +f) DNBR: Critical heat flux ratio F top :area F ΔH : Radial output peak value T in : Coolant temperature at the inlet after the abnormality T in0 : Average coolant temperature at the inlet before the abnormality a,b,c,d,e,f:Constant It is expressed as:

[0052] According to this configuration, it is possible to calculate an accurate estimate of the minimum DNBR using an evaluation model that is a regression equation.

[0053] The analysis method according to the third aspect is an analysis method executed by an analysis system 30 that performs an analysis of safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility, and the analysis system 30 estimates a minimum critical heat flux ratio, which is a parameter of the safety assessment, using an evaluation model. The evaluation model is a function of the critical heat flux ratio and the area F of a predetermined section in the axial power distribution of the fuel rods of the fuel assembly 20 loaded in the reactor core.top and the coolant temperature T at the coolant inlet side of the fuel assembly 20. in , T in0 and the radial power peak value F in the fuel assembly 20 ΔH The model represents the correlation with at least one of the areas F top and the coolant temperature T in , T in0 and the radial output peak value F ΔH and a step S2 of acquiring at least one of the area F top and the coolant temperature T in , T in0 and the radial output peak value F ΔH and Step S3 of calculating the minimum critical heat flux ratio at the time of the abnormality from the evaluation model based on at least one of the above.

[0054] According to this configuration, by inputting input parameters including the area and at least one of the coolant temperature and the radial power peak value into the evaluation model, it is possible to calculate an accurate estimate of the minimum DNBR.

[0055] The analysis program P according to the fourth aspect is an analysis program P executed by an analysis system 30 that performs an analysis of safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility. The analysis system 30 estimates a minimum critical heat flux ratio, which is a parameter of the safety assessment, using an evaluation model. The evaluation model is a function of the critical heat flux ratio and the area F of a predetermined section in the axial power distribution of the fuel rods of the fuel assembly 20 loaded in the reactor core. top and the coolant temperature T at the coolant inlet side of the fuel assembly 20. in , T in0 and the radial power peak value F in the fuel assembly 20 ΔH The model represents the correlation with at least one of the areas F top and the coolant temperature T in , T in0 and the radial output peak value F ΔH and a step S2 of acquiring at least one of the area F top and the coolant temperature Tin , T in0 and the radial output peak value F ΔH and Step S3 of calculating the minimum critical heat flux ratio at the time of the abnormality from the evaluation model based on at least one of the above.

[0056] According to this configuration, by inputting input parameters including the area and at least one of the coolant temperature and the radial power peak value into the evaluation model, it is possible to calculate an accurate estimate of the minimum DNBR. [Explanation of symbols]

[0057] 30 Analysis System 31 Arithmetic section 32 Storage section P analysis program

Claims

1. An analysis system for analyzing safety assessments in the event of an abnormality in a main steam pipe installed in a nuclear facility, a calculation unit that estimates a minimum critical heat flux ratio, which is a parameter for safety assessment, using an assessment model; the evaluation model is a model that represents a correlation between a critical heat flux ratio, an area of ​​a predetermined section in an axial power distribution of a fuel rod of a fuel assembly loaded in a reactor core, a coolant temperature at a coolant inlet side of the fuel assembly, and a radial power peak value of the fuel assembly, The calculation unit acquiring the area, the coolant temperature, and the radial power peak value; calculating the minimum critical heat flux ratio in the abnormal state from the evaluation model based on the area, the coolant temperature, and the radial power peak value; The evaluation model is represented by a regression equation, The regression equation is Estimated DNBR = (a·F top + b) (c·F ΔH + d) (e·T in / T in0 + f) DNBR: Critical heat flux ratio F top: area F ΔH: Radial output peak value T in : Coolant temperature at the inlet after the abnormality T in0 : Average coolant temperature at the inlet before the abnormality a, b, c, d, e, f: constant The analytical system represented by

2. An analysis method executed by an analysis system that analyzes safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility, comprising: The analysis system estimates a minimum critical heat flux ratio, which is a parameter for safety assessment, using an assessment model; the evaluation model is a model that represents a correlation between a critical heat flux ratio, an area of ​​a predetermined section in an axial power distribution of a fuel rod of a fuel assembly loaded in a reactor core, a coolant temperature at a coolant inlet side of the fuel assembly, and a radial power peak value of the fuel assembly, acquiring the area, the coolant temperature, and the radial power peak value; calculating the minimum critical heat flux ratio in the abnormal state from the evaluation model based on the area, the coolant temperature, and the radial power peak value; The evaluation model is represented by a regression equation, The regression equation is Estimated DNBR = (a·F top + b) (c·F ΔH + d) (e·T in / T in0 + f) DNBR: Critical heat flux ratio F top: area F ΔH: Radial output peak value T in : Coolant temperature at the inlet after the abnormality T in0 : Average coolant temperature at the inlet before the abnormality a, b, c, d, e, f: constant The analytical method is expressed as:

3. An analysis program executed in an analysis system that analyzes safety assessment in the event of an abnormality in a main steam pipe installed in a nuclear facility, The analysis system estimates a minimum critical heat flux ratio, which is a parameter for safety assessment, using an assessment model; the evaluation model is a model that represents a correlation between a critical heat flux ratio, an area of ​​a predetermined section in an axial power distribution of a fuel rod of a fuel assembly loaded in a reactor core, a coolant temperature at a coolant inlet side of the fuel assembly, and a radial power peak value of the fuel assembly, acquiring the area, the coolant temperature, and the radial power peak value; calculating the minimum critical heat flux ratio in the abnormal state from the evaluation model based on the area, the coolant temperature, and the radial power peak value; The evaluation model is represented by a regression equation, The regression equation is Estimated DNBR = (a·F top + b) (c·F ΔH + d) (e·T in / T in0 + f) DNBR: Critical heat flux ratio F top: area F ΔH: Radial output peak value T in : Coolant temperature at the inlet after the abnormality T in0 : Average coolant temperature at the inlet before the abnormality a, b, c, d, e, f: constant An analysis program represented by.

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