JP7117207B2 - Reactor Evaluation System, Reactor Evaluation Method, and Reactor Evaluation Program - Google Patents

Description

TECHNICAL FIELD The present invention relates to a nuclear reactor evaluation apparatus, a nuclear reactor evaluation method, and a nuclear reactor evaluation program.

In order to safely design nuclear reactors such as light water reactors, there is a method for predicting and evaluating the critical heat flux, which is the local thermal power leading to detachment from nucleate boiling, using an improved statistical thermal design method. known (see, for example, Non-Patent Document 1).

"Mitsubishi Public Document MHI-NES-1009 Revision 2 'Improved Statistical Thermal Design Method'", December 2009

The method of Non-Patent Document 1 deterministically considers the bending penalty, which is a parameter that represents the influence of the bending of the fuel rods in the core of the nuclear reactor on the critical heat flux and the critical heat flux ratio, which is the value of the ratio. is doing. Specifically, in the method of Non-Patent Document 1, based on the test analysis results of the critical heat flux, for each predetermined internal pressure in the reactor, the bending penalty and the high-temperature heating rod average heat flux, that is, the reactor By applying the least squares method approximation to the correlation with the average value of the heat flux of the fuel rods internally generating heat at a high temperature, the bending penalty is calculated as a linear expression of the average heat flux of the high-temperature heating rods.

Here, in the method of Non-Patent Document 1, by using the above-described linear expression of the high-temperature heating rod average heat flux of the bending penalty and combining the core cooling conditions of the reactor that conservatively give the severest result, bending There was room for rationalization because the penalty was calculated based on a more conservative framework evaluation than the actual situation.

The present invention has been made in view of the above. An object of the present invention is to provide a nuclear reactor evaluation apparatus, a nuclear reactor evaluation method, and a nuclear reactor evaluation program that enable evaluation.

In order to solve the above-described problems and achieve the object, a nuclear reactor evaluation apparatus includes an evaluation unit that evaluates the critical heat flux ratio of a nuclear reactor, and the evaluation unit includes a heat-flux ratio in the core of the nuclear reactor. By performing analysis, a local quality, which is a parameter representing the quality of thermal equilibrium, is calculated for each divided region divided into a plurality of regions in the core, and based on the local quality, a local critical heat flux ratio Determining a local space where the local critical heat flux ratio is the minimum, and based on the local quality in the local space and the amount of bending of the fuel rod in the reactor, the bending of the fuel rod is determined to be the critical heat flux ratio. A bending penalty, which is a parameter representing the influence, is calculated, and the calculated bending penalty is linked to the critical heat flux ratio of the nuclear reactor calculated by the thermal-hydraulic analysis.

According to this configuration, the bending penalty is calculated based on the local quality, which is the quality of thermal equilibrium in the local space, and the bending amount of the fuel rod. A reactor can be evaluated after making a rational evaluation according to the actual situation.

In this configuration, the evaluation unit calculates the bending penalty using a bending penalty correlation formula having the local quality and the bending amount of the fuel rod as input parameters, and the bending penalty correlation formula is As the local quality, a value obtained by subtracting the saturated enthalpy of the liquid phase of the coolant from the enthalpy of the coolant with respect to the latent heat of the coolant is used. If the local quality is greater than the first threshold and less than or equal to the second threshold, the local quality is calculated using a function that monotonically decreases from the predetermined positive number according to the local quality, and the local If the quality is greater than the second threshold, it is preferably calculated as 0. According to this configuration, the bending penalty can be evaluated rationally according to the actual situation.

In the configuration using the value obtained by subtracting the saturated enthalpy of the liquid phase of the coolant from the enthalpy of the coolant with respect to the latent heat of the coolant as the local quality, the amount of bending of the fuel rod is the fuel rod having the bending Let D be the distance between the adjacent fuel rods when the fuel rods are included, and D be the distance between the adjacent fuel rods when the bent fuel rods are not included. Then, 1-(D' /D), and the evaluation unit preferably determines the first threshold value of the local quality according to the amount of bending. With this configuration, the dependence of the curve penalty equation on the amount of bending of the fuel rod becomes easy to understand, so that the curve penalty can be handled with a simple equation for fuel rods having various degrees of curve. Here, the distance between adjacent fuel rods is the gap between adjacent fuel rods, that is, the distance between the outer peripheries of adjacent fuel rods. The distance between adjacent fuel rods is not limited to the above, and may be, for example, the distance between the central axes of adjacent fuel rods. Also, the amount of bending statistically treats the maximum value in the axial direction of the fuel rod.

In these configurations, it is preferable that the evaluation unit statistically treats the bending amount of the fuel rod as a random variable, and calculates the bending penalty using a bending penalty correlation formula. According to this configuration, it is possible to incorporate the bending penalty, which has been rationally evaluated according to the actual situation, into the evaluation and process it while maintaining the rationality and nature of the penalty.

Further, in order to solve the above-described problems and achieve the object, the nuclear reactor evaluation apparatus includes an evaluation unit that evaluates the critical heat flux ratio of the nuclear reactor, and the evaluation unit includes: A thermal hydraulic analysis is performed, and a penalty factor parameter related to a predetermined penalty is statistically treated as a random variable together with an input parameter handled in the thermal hydraulic analysis to calculate the predetermined penalty, and the calculated predetermined penalty is used as the It is characterized in that it is coupled to the critical heat flux ratio of the nuclear reactor calculated by thermal hydraulic analysis.

According to this configuration, when the penalty factor parameter, which is the factor parameter of the predetermined penalty, is known and the predetermined penalty can be calculated based on the penalty factor parameter, penalties other than the bending penalty can also be conservatively calculated. It is possible to evaluate the nuclear reactor after making a rational evaluation according to the actual situation, instead of making a framed evaluation.

In the configuration that combines the predetermined penalty, the evaluation unit may combine the calculated predetermined penalty and the uncertainty parameter with the critical heat flux ratio of the reactor calculated in the thermal hydraulic analysis. preferable. With this arrangement, a predetermined penalty can be tied to the critical heat flux ratio of the reactor in the same way as the uncertainty parameter.

In order to solve the above-mentioned problems and achieve the object, a nuclear reactor evaluation method performs a heat-hydraulic analysis in the core of a nuclear reactor to obtain a heat equilibrium for each of the plurality of divided regions in the core. a local quality calculating step of calculating a local quality which is a parameter representing the quality of the local space for determining a local space in which a local critical heat flux ratio, which is a local critical heat flux ratio, is minimized based on the local quality and calculating a bow penalty, a parameter representing the effect of the bending of the fuel rods on the critical heat flux ratio, based on the determining step and the local quality in the local space and the amount of bending of the fuel rods in the reactor. and a critical heat flux ratio evaluation step of combining the bending penalty calculated in the bending penalty calculating step with the critical heat flux ratio of the reactor calculated in the thermal hydraulic analysis. characterized by According to this configuration, as in the corresponding nuclear reactor evaluation apparatus described above, the bending penalty is calculated based on the local quality, which is the quality of thermal equilibrium in the local space, and the bending amount of the fuel rod. It is possible to evaluate the nuclear reactor after making a rational evaluation according to the actual state of the reactor, rather than performing a conservative framework evaluation.

In this configuration, it is preferable to treat the bending amount of the fuel rod statistically as a random variable and calculate the bending penalty using a bending penalty correlation formula. According to this configuration, similarly to the above-described corresponding nuclear reactor evaluation apparatus, the bending penalty, which is rationally evaluated according to the actual situation, is incorporated into the evaluation and processed while maintaining the rationality and nature of the penalty. can do.

Further, in order to solve the above-described problems and achieve the object, a nuclear reactor evaluation method evaluates a critical heat flux ratio of a nuclear reactor, and comprises heat-flux in the core of the nuclear reactor. a penalty calculation step of calculating the predetermined penalty by performing an analysis and statistically treating the penalty factor parameter related to the predetermined penalty as a random variable together with the input parameters handled in the thermal fluid analysis; and the penalty calculation step. and a critical heat flux ratio evaluation step of combining the predetermined penalty calculated by the thermal hydraulic analysis with the critical heat flux ratio of the nuclear reactor calculated by the thermal hydraulic analysis. According to this configuration, as in the corresponding nuclear reactor evaluation apparatus described above, when the penalty factor parameter, which is the factor parameter of the predetermined penalty, is known, and the predetermined penalty can be calculated based on the penalty factor parameter For penalties other than bending penalties, the nuclear reactor can be evaluated after rational evaluation according to the actual situation, instead of conservative frame evaluation.

In the configuration in which the predetermined penalty is combined, in the critical heat flux ratio evaluation step, together with the calculated predetermined penalty, an uncertainty parameter is applied to the critical heat flux ratio of the reactor calculated in the thermal hydraulic analysis. A combination is preferred. With this arrangement, a predetermined penalty can be tied to the critical heat flux ratio of the reactor in the same manner as the uncertainty parameter, similar to the corresponding reactor evaluator described above.

In order to solve the above-mentioned problems and achieve the purpose, the nuclear reactor evaluation program has an evaluation unit that evaluates the critical heat flux ratio of the nuclear reactor by performing a heat-hydraulic analysis in the core of the nuclear reactor. A local quality calculation step of calculating a local quality, which is a parameter representing the quality of thermal equilibrium, for each divided region divided into a plurality of regions in the core; and a local limit, which is a local limit heat flux ratio, based on the local quality. a local space determination step of determining a local space with a minimum heat flux ratio; a bending penalty calculating step for calculating a bending penalty, which is a parameter representing an effect on the flux ratio; and a critical heat flux ratio evaluation step coupled to . According to this configuration, similarly to the corresponding reactor evaluation apparatus and reactor evaluation method described above, the bending penalty is calculated based on the local quality, which is the quality of thermal equilibrium in the local space, and the bending amount of the fuel rod. Therefore, it is possible to evaluate the reactor after making a rational evaluation according to the actual state of the reactor, instead of performing a conservative framed evaluation of the bending penalty.

In this configuration, it is preferable that the evaluation unit statistically treats the bending amount of the fuel rod as a random variable and calculates the bending penalty using a bending penalty correlation formula. According to this configuration, similarly to the above-described corresponding reactor evaluation apparatus and reactor evaluation method, the bending penalty, which is rationally evaluated according to the actual situation, while maintaining its rationality and nature as a penalty, It can be imported and processed in a heat-hydraulic analysis.

In addition, in order to solve the above-mentioned problems and achieve the purpose, the nuclear reactor evaluation program has an evaluation unit that evaluates the critical heat flux ratio of the nuclear reactor, performs a heat-hydraulic analysis in the core of the nuclear reactor, A penalty calculation step of calculating the predetermined penalty by statistically treating a penalty factor parameter related to the predetermined penalty as a random variable together with the input parameters handled in the heat flow analysis, and the predetermined penalty calculated in the penalty calculation step and a critical heat flux ratio evaluation step of combining the penalty with the critical heat flux ratio of the nuclear reactor calculated by the thermal hydraulic analysis. According to this configuration, as in the corresponding reactor evaluation apparatus and reactor evaluation method described above, the penalty factor parameter, which is the factor parameter of the predetermined penalty, is known, and the predetermined penalty is calculated based on the penalty factor parameter. If it is possible to do so, penalties other than bending penalties can be evaluated rationally according to the actual situation, instead of conservatively framed evaluation.

In the configuration that combines the predetermined penalties, in the critical heat flux ratio evaluation step, the calculated predetermined penalty and the uncertainty parameter of the nuclear reactor calculated in the heat-hydraulic analysis are added to the evaluation unit. It is preferred to couple it to a limiting heat flux ratio. According to this configuration, similar to the corresponding reactor evaluation apparatus and method described above, a predetermined penalty can be combined with the critical heat flux ratio of the reactor in the same manner as the uncertainty parameter. can.

FIG. 1 is a schematic diagram showing an example of a nuclear reactor evaluation system and a nuclear reactor in which the nuclear reactor evaluation system is used according to an embodiment of the present invention. FIG. 2 is a configuration diagram showing an example of a detailed configuration of the core of FIG. 1. In FIG. FIG. 3 is a configuration diagram showing an example of a detailed flow channel configuration of the fuel assembly of FIG. FIG. 4 is a flow chart of a nuclear reactor evaluation method according to an embodiment of the present invention. FIG. 5 is a diagram showing the handling of critical heat flux ratio data in the nuclear reactor evaluation method of FIG. FIG. 6 is an explanatory diagram for explaining each step of the flowchart of FIG. 4 in handling the data of FIG. FIG. 7 is a configuration diagram showing an example of the distance between adjacent fuel rods in the fuel assembly of FIG. FIG. 8 is a graph showing the correlation between bending penalty and local quality calculated by the reactor evaluator of FIG. FIG. 9 is a graph showing the correlation between the bending penalty and the high temperature heating bar average heat flux calculated by the conventional method. FIG. 10 is a diagram showing a comparison between the handling of critical heat flux ratio data by the nuclear reactor evaluation system 10 of FIG. 1 and the handling of critical heat flux ratio data by a conventional method.

EMBODIMENT OF THE INVENTION Below, embodiment which concerns on this invention is described in detail based on drawing. In addition, this invention is not limited by this embodiment. In addition, components in the embodiments include those that can be easily replaced by those skilled in the art, or those that are substantially the same. Furthermore, the components described below can be combined as appropriate.

[Embodiment]
FIG. 1 is a schematic diagram showing a nuclear reactor evaluation system 10 according to an embodiment of the present invention and a nuclear reactor 20 as an example of a nuclear reactor in which the nuclear reactor evaluation system 10 is used. The reactor evaluation system 10 includes an evaluation unit 12 as shown in FIG. The evaluation unit 12 is electrically connected to each part of the nuclear reactor 20 so as to be able to communicate information, and acquires various parameters such as design values related to each part from each part of the nuclear reactor 20 . In the present embodiment, the evaluation unit 12 is configured to directly acquire various parameters such as measured values relating to each unit from each unit of the nuclear reactor 20. and a form of acquiring these various parameters via a storage medium or storage device exemplified by flash memory and the like.

The evaluation unit 12 includes an input unit, a storage unit, and a processing unit. The evaluation unit 12 is exemplified by a computer. The input unit is, for example, a touch panel integrated with an interface that accepts input of environmental data from the reactor 20, which is an externally connected device, and a mouse, keyboard, display device, etc. that is an interface that accepts input from the user. It transmits information received as input to the storage unit or the processing unit. The storage unit has a storage medium or storage device such as RAM, ROM, and flash memory, and stores a reactor evaluation program, which is a software program processed by the processing unit, and data referred to by this software program. do. The storage unit also functions as a storage area in which the processing unit temporarily stores processing results and the like. The processing unit reads a software program or the like from the storage unit and processes the function according to the contents of the software program. It exhibits the function of implementing the reactor evaluation method according to the embodiment. The processing unit can display information read from the storage unit, processed information, and the like on a display device or the like connected to the evaluation unit 12 .

The nuclear reactor 20 includes a core 30 inside, as shown in FIG. The core 30 is filled with a flowing coolant 25 as shown in FIG. Inside the nuclear reactor 20, the coolant 25 is a medium for taking out the thermal energy obtained by the nuclear fission of the nuclear fuel contained in the core 30 to the outside of the nuclear reactor 20. Also works as material.

FIG. 2 is a configuration diagram showing an example of the detailed configuration of the core 30 of FIG. 1. As shown in FIG. FIG. 3 is a configuration diagram showing an example of the detailed configuration of the fuel assembly 32 of FIG. 2. As shown in FIG. FIG. 2 shows only the lower right quarter portion of FIG. 2 for the core 30 of FIG. FIG. 3 shows only the lower right quarter portion in FIG. 3 of the fuel assembly 32 of FIG.

As shown in FIG. 2, the core 30 includes a plurality of arranged fuel assemblies 32 and control rods (not shown) that can be inserted into and removed from the fuel assemblies 32 . As shown in FIG. 3, the fuel assembly 32 is formed by arranging a plurality of fuel rods 34 containing nuclear fuel and processed into a rod-like composite. The fuel assembly 32 further includes a control rod guide thimble 36 into which a control rod for controlling nuclear fission of the nuclear fuel contained in the fuel assembly 32 can be inserted and removed, and various parameters such as measurements related to each part of the core 30. and an instrumentation guide thimble 38 for inserting the meter. As shown in FIG. 3, the fuel assembly 32 has a plurality of regions 40 surrounded by fuel rods 34 arranged therein. Note that the region 40 includes not only the region surrounded by the four fuel rods 34 in FIG.

The reactor 20 is exemplified by a pressurized water reactor (PWR) in this embodiment, but the present invention is not limited to this and may be other types of reactors. If the reactor 20 is a pressurized water reactor, it comprises one or more steam generators (SG). In this case, the reactor 20 constitutes, together with these steam generators, a primary cooling system (Reactor Coolant System, RCS), which is a closed system in which a coolant 25 as a primary coolant flows. The heat energy taken out of the reactor 20 by the coolant 25 is transmitted to the secondary cooling system in the steam generator and contributes to nuclear power generation.

In order to safely design such a nuclear reactor 20, the evaluation unit 12 uses a generalized statistical thermal-design method (GSTM) to estimate the local heat that leads to detachment from nucleate boiling. Predicting the critical heat flux (Departure from Nucleate Boiling, DNB) that is the output and the critical heat flux ratio (Departure from Nucleate Boiling Ratio, DNBR) that is the value of its ratio, especially the minimum limit that takes the smallest value among them The heat flux ratio (minimum DNBR) is extracted and evaluated.

The operation of the nuclear reactor evaluation system 10 according to the embodiment having the configuration as described above will be described below. FIG. 4 is a flow chart of a nuclear reactor evaluation method according to an embodiment of the present invention. A nuclear reactor evaluation method according to an embodiment of the present invention is an example of the operation of the nuclear reactor evaluation apparatus 10 . In the nuclear reactor evaluation method according to the embodiment of the present invention, the evaluation unit 12 performs calculation using the Monte Carlo method, that is, Monte Carlo calculation, as described below with reference to FIG. Predict and evaluate the flux ratio and evaluate the design limits for the critical heat flux ratio. As shown in FIG. 4, the nuclear reactor evaluation method according to the embodiment includes a core coolant condition setting step S1, an input parameter setting step S2, a minimum critical heat flux ratio calculation step S3, a bending penalty calculation step S4, It includes a critical heat flux ratio evaluation step S5, a calculation count determination step S6, a probability distribution condition determination step S7, and a core coolant condition change step S8.

FIG. 5 is a diagram showing the handling of critical heat flux ratio data in the nuclear reactor evaluation method of FIG. In the nuclear reactor evaluation method of FIG. 4, as shown in FIG. The reactor pressure 53, the primary coolant mean temperature 54, the reactor power 55, the engineering heat channel coefficients for enthalpy rise 56, and the nuclear enthalpy rise heat channel coefficients 57 are defined by the uncertainties of these input parameters. are collectively considered together, and furthermore, the bending amount 58 of the fuel rod 34, which is an important parameter for the bending penalty 68 (see FIG. 6) that was rationally evaluated according to the actual situation, and the critical heat flux test Using the uncertainty of the ratio 59 of the predicted value (P) to the measured value (M) of the critical heat flux based on the critical heat flux correlation formula (DNB correlation formula) created based on the data, the predetermined critical heat flux By obtaining the probability distribution 60 of the random variable (coupling probability Z) regarding the ratio, the critical heat flux ratio is evaluated, and the design limit value of the critical heat flux ratio is evaluated.

Here, primary coolant flow rate 51, core bypass flow rate 52, reactor pressure 53, primary coolant average temperature 54, reactor power 55, engineering heat channel coefficient for enthalpy rise 56, engineering heat for enthalpy rise Each of the channel coefficient 56, the amount of bending 58, and the ratio 59 of the predicted value (P) by the critical heat flux correlation formula to the measured value (M) of the critical heat flux outputs one value according to a random number. Treated statistically as a random variable.

As shown in FIG. 5, the lower limit value 61 of the critical heat flux ratio at which the probability distribution 60 has a predetermined probability, such as a probability of 95%, is equal to 1.0. Also, the permissible limit value 62 of the critical heat flux ratio can be calculated based on the probability distribution 60 . Note that the permissible limit value 62 of the critical heat flux ratio is appropriately set for each plant. Also, the normal value 63 of the critical heat flux ratio is the value of the critical heat flux ratio when the nuclear reactor 20 is normally operated.

As shown in FIG. 5, the difference 64 between the design limit value of the critical heat flux ratio and the allowable limit value 62 of the critical heat flux ratio in the probability distribution 60 is the critical heat flow rate allotted to the penalty for conservative framing evaluation. It represents the margin of the flux ratio. Also, the difference 65 between the allowable limit value 62 of the critical heat flux ratio and the normal value 63 of the critical heat flux ratio in the probability distribution 60 is the critical heat flux ratio relative to the allowable limit when the reactor 20 is operated under normal conditions. This indicates a design margin. The evaluation unit 12 of the reactor evaluation apparatus 10 can evaluate that the reactor 20 is extremely safe during normal operation according to the magnitude of the difference 65 .

FIG. 6 is an explanatory diagram for explaining each step of the flowchart of FIG. 4 in handling the data of FIG. The core coolant condition setting step S1 corresponds to the processing 71 shown in FIG. is a step of setting

The input parameter setting step S2 corresponds to the processing 72 shown in FIG. Below is the step of setting each input parameter to be treated statistically by generating random numbers. Specifically, in the input parameter setting step S2, a random number is generated for each input parameter, and each input parameter is set based on the generated random number.

In the input parameter setting step S2, in this embodiment, for example, as shown in FIG. , the engineering heat channel coefficient 56 and the nuclear enthalpy rise heat channel coefficient 57 related to enthalpy rise are set as input parameters that are statistically handled by generating random numbers, but the present invention is limited to this. Instead, each input parameter can be set as appropriate.

In the input parameter setting step S2, in the present embodiment, the evaluation unit 12 also determines the bending amount 58 of the fuel rod 34 by A random number is generated, and the bending amount 58 of the fuel rod 34 is set based on the generated random number.

The bending amount 58 of the fuel rod 34 used in the nuclear reactor evaluation method according to the embodiment of the present invention will be described with reference to FIG. FIG. 7 is a configuration diagram showing an example of the distance between adjacent fuel rods 34 of the fuel assembly 32 of FIG. In the nuclear reactor evaluation method according to the embodiment of the present invention, as shown in FIG. Assuming that the distance between the adjacent fuel rods 34 without the fuel rods 34 is D, the bending amount 58 of the fuel rods 34 is preferably 1-(D'/D). Here, the distance between the adjacent fuel rods 34 is the gap between the adjacent fuel rods 34, that is, the distance between the outer peripheries of the adjacent fuel rods 34 as shown in FIG. When such a value is used as the bending amount 58 of the fuel rod 34, for example, in the case of so-called contact bending, in which a bent fuel rod 34 is in contact with an adjacent fuel rod 34, the bent fuel rod 34 is Since the distance D' between them is 0, the bending amount 58 is 1, and when expressed in percentage [unit: %], it is 100%. Such a bending amount 58 expresses the effect on the flow rate of the coolant 25 between the fuel rods 34 in an easy-to-understand manner. The distance between the adjacent fuel rods 34 to be used for the bending amount 58 of the fuel rods 34 is not limited to the above, and may be, for example, the distance between the central axes of the adjacent fuel rods 34. . Also, the bending amount 58 of the fuel rod 34 statistically treats the maximum value in the axial direction of the fuel rod 34 .

In the input parameter setting step S2, in this embodiment, the evaluation unit 12 further uses the uncertainty parameter of the ratio 59 of the predicted value (P) to the measured value (M) of the critical heat flux that can be set by the critical heat flux correlation formula. Similarly to the execution for each input parameter and the bending amount 58 of the fuel rod 34, a random number is generated for the uncertainty parameter, and the uncertainty parameter is set based on the generated random number do. Here, in the nuclear reactor evaluation method according to the embodiment of the present invention, the uncertainty parameter of the predicted value (P) for the measured critical heat flux (M) is obtained by dividing the measured critical heat flux by the predicted value. It is preferred to use the resulting quotient value (M/P).

The minimum limit heat flux ratio calculation step S3 corresponds to the processing 73 shown in FIG. 6 in terms of data handling. A step of calculating a minimum critical heat flux ratio 66, which is the minimum value of the critical heat flux ratios, based on each input parameter except for and. The minimum critical heat flux ratio 66 calculated in the minimum critical heat flux ratio calculation step S3 takes into consideration the uncertainty of each input parameter, and also the bending penalty 68 calculated based on the bending amount 58 of the fuel rod 34. It was calculated excluding any penalties including and uncertainty parameters.

In the minimum critical heat flux ratio calculation step S<b>3 , in this embodiment, the evaluation unit 12 performs thermal fluid analysis in the core 30 of the nuclear reactor 20 when calculating the minimum critical heat flux ratio 66 . In the minimum critical heat flux ratio calculation step S3, taking advantage of this, together with the calculation of the minimum critical heat flux ratio 66, the heat equilibrium quality is calculated for each sub-region (divided region) divided into a plurality of regions in the core 30. A local quality calculation step is executed to calculate a local quality 67 which is a parameter representing . Specifically, in the local quality calculation step in the minimum limit heat flux ratio calculation step S3, the evaluation unit 12 solves the three-dimensional flow of heat flow by performing a heat-hydraulic analysis, and calculates the overall three-dimensional flow in the core 30 The local quality 67 is calculated together with the values of the pressure, flow rate, temperature, etc. of the coolant 25 for each mesh-shaped small region cut from the figure.

Specifically, in the local quality calculation step in the minimum limit heat flux ratio calculation step S3, it is preferable that the evaluation unit 12 calculates the thermal equilibrium quality x shown in (Equation 1) below for each of the small regions described above.

Here, the thermal equilibrium quality x on the left side of the above (Equation 1) is a dimensionless parameter. The denominator of the fractional expression on the right side of the above (Equation 1) is the difference between the vapor saturation enthalpy _hg [unit: J/kg] of the coolant 25 and the liquid phase saturation enthalpy _hf [unit: J/kg]. It represents a certain latent heat h _fg [unit: J/kg]. The numerator of the fractional expression on the right side of the above (Equation 1) is obtained by subtracting the saturation enthalpy h _f [unit: J/kg] of the liquid phase of the coolant 25 from the enthalpy h _m [unit: J/kg] of the coolant 25. is a value.

In the minimum critical heat flux ratio calculation step S3, the evaluation unit 12 further determines the local heat flux ratio, which is the local critical heat flux ratio, based on the local quality 67 calculated in the local quality calculation step. Perform a local space determination step to determine the space. Specifically, in the local space determination step in the minimum critical heat flux ratio calculation step S3, the evaluation unit 12 determines a small region having the minimum local critical heat flux ratio as the local space.

In the minimum critical heat flux ratio calculation step S3, the calculation of the minimum critical heat flux ratio 66, the calculation of the local quality 67, and the determination of the local space are executed in the core 30 of the series of nuclear reactors 20. In the analysis, it is executed in conjunction.

The bending penalty calculation step S4 corresponds to the processing 74 shown in FIG. 6 in terms of data handling. Based on the local quality 67 calculated in the local quality calculation step in the local space determined in the local space determination step in S3, a bending penalty correlation formula with input parameters of the bending amount 58 and the local quality 67 of the fuel rod 34 is used. The next step is to calculate a bending penalty 68, which is a parameter representing the effect of the bending of the fuel rods 34 on the critical heat flux ratio.

FIG. 8 is a graph 80 showing the correlation between bend penalty 68 and local quality 67 calculated by reactor evaluator 10 of FIG. A graph 80 graphically represents the bending penalty correlation formula for a given amount of bending 58 of the fuel rod 34. As shown in FIG. 68 [units; dimensionless] and has a first relationship curve 81 and a second relationship curve 87 .

The first relationship curve 81 is a graph 80 showing a function used by the evaluation unit 12 in calculating the bending penalty 68 when the bending amount 58 of the fuel rod 34 is 100% in the bending penalty calculation step S4. is. The first relationship curve 81 has a straight line 84 , a straight line 85 and a straight line 86 bordering on a first threshold 82 and a second threshold 83 of the local quality 67 .

A straight line 84 is based on a function that calculates the bending penalty 68 as a predetermined positive number when the local quality 67 is less than or equal to the first threshold 82, and is shown in graph 80 as a straight line parallel to the horizontal axis. . A straight line 85 is a function for calculating the bending penalty 68 when the local quality 67 is greater than the first threshold 82 and less than or equal to the second threshold 83. , which is connected with a straight line 84 where the local quality 67 is at the first threshold 82 and with a straight line 86 where the local quality 67 is at the second threshold 83 as Shown in graph 80 . A straight line 86 is a function that calculates the bend penalty 68 at 0 if the local quality 67 is greater than the second threshold 83, and is shown in graph 80 as a straight line superimposed on the horizontal axis.

The second relationship curve 87 is a graph 80 showing a function used by the evaluation unit 12 in calculating the bending penalty 68 when the bending amount 58 of the fuel rod 34 is less than 100% in the bending penalty calculation step S4. It is. Here, as a case where the bending amount 58 of the fuel rod 34 is less than 100%, the case where the bending amount 58 of the fuel rod 34 is 85% is exemplified in this embodiment, but the present invention is not limited to this. , the amount of bending 58 of any fuel rod 34 less than 100%. A second relationship curve 87 has a portion of the straight line 84 , a straight line 89 and a straight line 86 bounded by a first threshold 88 and a second threshold 83 of the local quality 67 .

The straight line 84 is common between the first relationship curve 81 and the second relationship curve 87 in the region below the first threshold value 88 . That is, the evaluation unit 12 calculates the same bending penalty 68 without depending on the bending amount 58 of the fuel rod 34 under the condition that the local quality 67 is extremely low.

A straight line 89 is a function for calculating the bending penalty 68 when the local quality 67 is greater than the first threshold 88 and less than or equal to the second threshold 83, and is a function that monotonically decreases from a predetermined positive number according to the local quality 67. is based on and is connected with straight line 84 where local quality 67 is at first threshold 88 and with straight line 86 where local quality 67 is at second threshold 83 as a downward sloping straight line Shown in graph 80 .

Also, the second threshold value 83 and the straight line 86 are common between the first relationship curve 81 and the second relationship curve 87 . That is, the evaluation unit 12 calculates the bending penalty 68 as 0 without depending on the bending amount 58 of the fuel rod 34 under the condition that the local quality 67 is equal to or higher than a certain value.

Here, the first threshold 88, which is one of the boundaries of the second relationship curve 87, is a value smaller than the first threshold 82, which is one of the boundaries of the first relationship curve 81. The small degree can be appropriately determined according to the value of the bending amount 58 of the fuel rod 34 . That is, the first threshold is represented by a function that monotonically increases according to the value of the amount of bending 58 and reaches the first threshold 82 when the amount of bending 58 reaches 100%. Therefore, as the bending amount 58 of the fuel rod 34 increases, the evaluation unit 12 calculates the bending penalty 68 as a predetermined positive number, which is the maximum value, to the extent that the local quality 67 increases.

Further, from the relationship between the first threshold 82 and the first threshold 88 and the relationship between the second threshold 83, the magnitude of the slope of the straight line 85 in the first relationship curve 81 is It is larger than the slope of the straight line 89 . Furthermore, the slope of the straight line representing the region larger than the threshold monotonically increases according to the bending amount 58 of the fuel rod 34 . Therefore, the bending penalty 68 is adjusted so that the evaluation unit 12 tends to increase the bending penalty 68 more sensitively in accordance with the decrease from the second threshold value 83 as the bending amount 58 of the fuel rod 34 increases. Calculate

As described above, the relationship between the bending penalty 68 and the local quality 67 has various tendencies depending on the bending amount 58 of the fuel rods 34 . Therefore, by utilizing this relationship, the evaluation unit 12 calculates the bending penalty 68 for the bending amount 58 of the fuel rod 34 for which the relationship between the bending penalty 68 and the local quality 67 is not prepared. , depending on the bending amount 58 of the fuel rod 34, the threshold of the local quality 67 can be determined, the relationship between the bending penalty 68 and the local quality 67 can be determined, and the bending penalty 68 can be calculated as appropriate.

In this embodiment, for a region where the local quality 67 is greater than the threshold, the relationship between the bend penalty 68 and the local quality 67 is represented by a straight line in which the bend penalty 68 monotonously decreases according to the local quality 67. Although shown, the present invention is not limited to this, and if the curve penalty 68 is monotonically decreasing with respect to the local quality 67, it may be represented by a curve.

FIG. 9 is graphs 110, 120 showing the correlation between the bending penalty and the hot-heat bar average heat flux calculated by conventional methods. Graphs 110 and 120 shown in FIG. 9 are comparative examples of the graph 80 shown in FIG. Here, the hot-heat rod average heat flux refers to the average value of the heat flux of the fuel rods that generate heat at a high temperature inside the nuclear reactor. Graph 110 shown in FIG. 9A is for the case where the pressure inside the reactor is 14.5 MPa, and graph 120 shown in FIG. .6 MPa.

The graph of the correlation between the bending penalty and the high-temperature heating rod average heat flux according to the comparative example is the graph 110 shown in FIG. As shown in the graph 120 shown in FIG. 9B when the internal pressure is 16.6 MPa, it differs depending on the internal pressure of the nuclear reactor. A plurality of them are created for each pressure.

In the graph 110, as shown in FIG. 9A, the horizontal axis is the high-temperature heating bar average heat flux [unit; MW/m ² ], and the vertical axis is the bending penalty [unit; dimensionless]. It has a group 112 , a second group of plots 114 , a first relationship curve 116 and a second relationship curve 118 .

Both the first group of plots 112 and the second group of plots 114 are for the case where the fuel rod has contact bending in the thimble cell, which is the region adjacent to the high-temperature heat rod and the control rod guide thimble, that is, when the bending amount is 100%. Fig. 110 is a group of points plotted on a graph 110 for a combination of bending penalty and hot bar average heat flux values for a given case; The first group of plots 112 and the second group of plots 114 differ in that the direction in which the contact bar is bent is based on different tests.

A first relational curve 116 represented by a solid line is obtained by fitting the bending penalty with a linear expression of the average heat flux of the high temperature heating bar by applying the least squares method approximation based on the first plot group 112 . The first relationship curve 116 was not significantly different from that fitted based on the second group of plots 114 in the same manner as the first relationship curve 116 .

A second relationship curve 118, represented by a dashed line, is fitted in the same manner as the first relationship curve 116 in the typical cell, the region adjacent to the hot rods and not adjacent to the control rod guide thimbles. It was obtained by The results show that the first relationship curve 116 for the thimble cell has a higher bending penalty than the second relationship curve 118 for the typical cell.

As shown in (B) of ^FIG . 9, the graph 120 is similar to the graph 110 shown in (A) of FIG. It is a curvature penalty [units; dimensionless] and has a first group of plots 122 , a second group of plots 124 , a first relationship curve 126 and a second relationship curve 128 . The first group of plots 122, the second group of plots 124, the first relationship curve 126 and the second relationship curve 128 in the graph 120 are all the first group of plots 112, the second group of plots 114 and the first relationship curve in the graph 110. 116 and the second relationship curve 118 are obtained by changing the pressure inside the reactor from 14.5 MPa to 16.6 MPa, so the first plot group 112 and the second plot group in the graph 110, respectively 114, the first relationship curve 116, and the second relationship curve 118, and therefore detailed description thereof will be omitted.

Comparing the graphs 110 and 120, the first relationship curve 116 has a greater slope of the bending penalty for the hot-heat bar average heat flux than the first relationship curve 126, and the second relationship curve 118 has a higher slope than the first relationship curve 126. The slope of the bending penalty for hot bar average heat flux is greater than curve 128 . That is, in the comparative example, there is a tendency that the bend penalty increases more sensitively in accordance with the increase in the high-temperature heating rod average heat flux by reducing the internal pressure of the nuclear reactor. In addition, in the range of the high-temperature heating bar average heat flux of 1.0 MW / m ² or more and 1.8 MW / m ² or less, the first relationship curve 126 has a larger bending penalty than the first relationship curve 116, and the second Relationship curve 128 has a higher bending penalty than second relationship curve 118 . In other words, the comparative example tends to increase the bending penalty by increasing the internal pressure of the reactor.

As described with reference to FIG. 9, in the graphs 110 and 120 according to the comparative example, the bending penalty is represented by the correlation with the high-temperature heating bar average heat flux, and local cooling such as the local quality 67 It is not represented by the correlation with material conditions. On the other hand, in the present embodiment, since the local quality 67 is used as a counterpart for taking correlation with the bend penalty 68, the evaluation unit 12 can make a rational evaluation of the bend penalty 68 according to the actual situation. became possible.

The critical heat flux ratio evaluation step S5 corresponds to the process 75 shown in FIG. This is the step of combining the uncertainty parameter set in step S2 and the bend penalty 68 calculated in the bend penalty calculation step S4.

Specifically, in the critical heat flux ratio evaluation step S5, the evaluation unit 12 multiplies the minimum critical heat flux ratio 66 by the uncertainty parameter and a value obtained by subtracting the bend penalty 68 from 1, A post-combination minimum critical heat flux ratio Z that combines the uncertainty parameter and the bending penalty 68 is obtained once.

The number-of-calculations determination step S6 corresponds to the processing 76 shown in FIG. This is the step of determining whether the number of times, for example, 10000 times in this embodiment, has been repeated. If it is determined in the number-of-calculations determining step S6 that the statistically significant predetermined number of times has not been executed, that is, if it is determined as No in the number-of-calculations determining step S6, the statistically significant predetermined number of times is determined in the number-of-calculations determining step S6. Returning to the input parameter setting step S2, each step from the input parameter setting step S2 to the limit heat flux ratio evaluation step S5 is repeatedly executed until the execution is reached.

When it is determined that the number of times of calculation is determined to be statistically significant predetermined times in the number of calculations determination step S6, that is, when it is determined as Yes in the number of times of calculation determination step S6, the evaluation unit 12 performs acquisition as shown in FIG. A probability distribution 69 of the connection probability Z (random variable for 10000 times) regarding the critical heat flux ratio is created based on the statistically significant minimum critical heat flux ratio after the predetermined number of times of coupling. In this case, the evaluation unit 12 does not apply to any of the design parameter, each input parameter set in the input parameter setting step S2, and the bending amount 58 of the fuel rod 34, and statistically does not handle the parameter, Conservative framing evaluation may be performed so that the evaluation result with respect to the critical heat flux ratio becomes severe, and, for example, a selected fixed value may be used as a penalty and processed. For example, the evaluation unit 12 can take in and process the mixture penalty due to mixture of different types of fuels by conservatively framing evaluation and using selected fixed values.

The probability distribution condition determination step S7 and the core coolant condition change step S8 correspond to the processing 77 shown in FIG. 6 in terms of data handling. The probability distribution condition determination step S7 is a step in which the evaluation unit 12 determines whether or not the probability distribution 69 created in the calculation count determination step S6 satisfies a predetermined condition. In the probability distribution condition determination step S7, in the present embodiment, the evaluation unit 12 determines whether the lower limit value of the critical heat flux ratio at which the probability distribution 69 has a predetermined probability, for example, a probability of 95%, is equal to 1.0. determine whether In the probability distribution condition determination step S7, if it is determined that the predetermined condition is not satisfied, that is, the lower limit value of the critical heat flux ratio in the probability distribution 69 does not match 1.0, in the probability distribution condition determination step S7 A determination of No is made, and the reactor evaluation method proceeds to the core coolant condition change step S8.

The core coolant condition change step S8 moves the lower limit of the critical heat flux ratio in the probability distribution 69 closer to 1.0. As shown, this is a step of changing the conditions of the core 30 and the coolant 25 set in the core coolant condition setting step S1 in the direction of lowering the lower limit of the critical heat flux ratio. After the core coolant condition changing step S8, the reactor evaluation method shifts to the input parameter setting step S2 again. A new probability distribution 69 is created by repeating each step up to the critical heat flux ratio evaluation step S5.

In the probability distribution condition determination step S7, a predetermined condition is satisfied, that is, the lower limit of the critical heat flux ratio in the probability distribution 69 is equal to 1.0, and the probability distribution 69 is in a state like the probability distribution 60. If it is determined that it is, it is determined as Yes in the probability distribution condition determination step S7, the probability distribution 60 is determined, the critical heat flux and the critical heat flux ratio are evaluated based on this, and the critical heat flux ratio is designed After evaluating the limit value, the reactor evaluation method according to the present embodiment is terminated.

In the nuclear reactor evaluation method according to the embodiment of the present invention, as a result of diligent studies by the inventors, the evaluation unit 12 determines the bending amount 58 and the bending penalty 68 of the fuel rods 34 in a non-deterministic manner as in the prior art. It has become possible to handle it in a rational way. Therefore, in the nuclear reactor evaluation method according to the embodiment of the present invention, the bending amount 58 of the fuel rod 34 is set based on the random numbers generated in the same manner as the input parameters in the input parameter setting step S2. A bending penalty 68 can be calculated based on the bending amount 58 of the fuel rod 34 in the penalty calculation step S4, and the bending penalty 68 can be combined with the minimum critical heat flux ratio 66 together with the uncertainty parameter in the critical heat flux ratio evaluation step S5. became possible.

In the nuclear reactor evaluation method according to the present invention, the evaluation unit 12 performs the same processing as the bending amount 58 of the fuel rod 34, which is the factor parameter of the bending penalty 68, for penalties other than the bending penalty 68. may be performed and coupled to the minimum critical heat flux ratio 66 . That is, in the nuclear reactor evaluation method according to the present invention, when the penalty factor parameter, which is the factor parameter of the predetermined penalty, is known and the predetermined penalty can be calculated based on the penalty factor parameter, the evaluation unit 12 A penalty factor parameter, which is a factor parameter of a predetermined penalty, is set based on a random number generated, a predetermined penalty is calculated in a penalty calculation step similar to the bending penalty calculation step S4, and a critical heat flux ratio evaluation step A predetermined penalty may be coupled to the minimum critical heat flux ratio 66 along with the uncertainty parameter at S5. In the nuclear reactor evaluation method according to the present invention, the evaluation unit 12 determines the predetermined penalty for which such penalty factor parameters are known, as well as the bending amount 58 and the bending penalty 68 of the fuel rod 34. It is possible to make a rational evaluation according to the actual situation, which has been handled with a conservative frame evaluation so that the evaluation result for the bundle ratio becomes strict.

FIG. 10 is a diagram showing a comparison between the handling of critical heat flux ratio data by the nuclear reactor evaluation system 10 of FIG. 1 and the handling of critical heat flux ratio data by a conventional method. A probability distribution 60 shown in (A) of FIG. 10 represents the handling of critical heat flux ratio data by the nuclear reactor evaluation system 10, and is the same as that shown in FIG. A probability distribution 160 shown in FIG. 10B represents the handling of critical heat flux ratio data by a conventional method.

Lower limit 161 in probability distribution 160 corresponds to lower limit 61 in probability distribution 60 . Also, the permissible limit value 162 of the critical heat flux ratio corresponds to the permissible limit value 62 of the critical heat flux ratio, can be calculated based on the probability distribution 160, and is appropriately set for each plant. Further, the normal value 163 of the critical heat flux ratio corresponds to the normal value 63 of the critical heat flux ratio, and is the value of the critical heat flux ratio when the nuclear reactor 20 is normally operated.

The difference 164a and the difference 164b between the design limit value of the critical heat flux ratio and the allowable limit value 162 of the critical heat flux ratio in the probability distribution 160 correspond to the difference 64 in the probability distribution 60 and the difference 64 in the probability distribution 60. Similarly, it represents the margin of the critical heat flux ratio which is allotted for the penalty of conservative framing evaluation. The difference 164a is the same as the difference 64, and the difference 164b is the critical heat flux ratio headroom that would be reserved for a conservative framing estimate of the bend penalty.

In addition, the difference 165 between the allowable limit value 162 of the critical heat flux ratio and the normal value 163 of the critical heat flux ratio in the probability distribution 160 corresponds to the difference 65 in the probability distribution 60, and the reactor 20 is operated under normal conditions. It shows the design margin of the critical heat flux ratio to the allowable limit when operating.

In the nuclear reactor evaluation method using the reactor evaluation apparatus 10 of FIG. Since the framing evaluation is performed, the probability distribution 60 has a larger spread than the probability distribution 160 by the rationally combining the bending penalty 68, and the difference 165 is larger than the difference 165. It has a large difference 65. Therefore, the nuclear reactor evaluation method using the nuclear reactor evaluation apparatus 10 of FIG. 1 can secure a larger margin for the critical heat flux ratio than the conventional method.

Since the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment have the configurations as described above, the bending penalty 68 is the quality of the heat balance in the local space. Since it is calculated based on the local quality 67 and the bending amount 58 of the fuel rod 34, the bending penalty 68 is not evaluated conservatively, but after making a rational evaluation according to the actual situation, An evaluation of reactor 20 may be made. That is, the reactor evaluation apparatus 10 according to the embodiment, the reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment perform a rational evaluation of the bending penalty 68 according to the actual situation, and analyze the thermal hydraulics. By combining with the critical heat flux ratio calculated in , it can be incorporated as the spread of the probability distribution 60 shown in FIGS. 5 and 10, and the difference 164b in the conventional method can be eliminated. making it possible.

In the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment, the evaluation unit 12 further uses the local quality 67 and the bending amount 58 of the fuel rod 34 as input parameters. The curvature penalty 68 is calculated using the curvature penalty correlation equation , and this curvature penalty correlation equation is defined as the local quality 67 from the enthalpy of the coolant 25 to the latent heat of the coolant 25 to the liquid phase of the coolant 25 Using the value obtained by subtracting the saturation enthalpy, if the local quality 67 is less than the first thresholds 82, 88, the bending penalty 68 is calculated as a predetermined positive number, and if the local quality 67 is greater than the first thresholds 82, 88 If it is less than the second threshold 83, it is calculated using a function that monotonically decreases according to the local quality 67 from a predetermined positive number, and if the local quality 67 is greater than the second threshold 83, it is calculated as 0. ing. Therefore, the reactor evaluation apparatus 10 according to the embodiment, the reactor evaluation method according to the embodiment, and the reactor evaluation program according to the embodiment can evaluate the bending penalty 68 more rationally according to the actual situation. can.

The nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment further include the Assuming that the distance between the adjacent fuel rods 34 is D' and the distance between the adjacent fuel rods 34 when the curved fuel rods 34 are not included is D, then 1-(D'/D). , and the evaluation unit 12 determines the first thresholds 82 and 88 of the local quality 67 according to the bending amount 58 . Therefore, the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment have the dependence of the equation indicating the bending penalty 68 on the bending amount 58 of the fuel rod 34. For clarity, the bow penalty 68 can be treated with a simple formula for fuel rods 34 with varying degrees of bow.

In the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment, the evaluation unit 12 statistically evaluates the bending amount 58 of the fuel rod 34 as a random variable. The bending penalty 68 is calculated using the handling and bending penalty correlation formula. For this reason, the reactor evaluation apparatus 10 according to the embodiment, the reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment use the bending penalty 68 that is rationally evaluated according to the actual situation. and can be incorporated into the evaluation and processed while retaining the property as a penalty.

In the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment, the evaluation unit 12 performs thermal-hydraulic analysis in the core 30 of the nuclear reactor 20, This predetermined penalty is calculated by statistically treating the penalty factor parameter related to the penalty of the reactor 20 as a random variable together with the input parameter handled in the thermal hydraulic analysis, and the calculated predetermined penalty is calculated in the thermal hydraulic analysis. It is characterized by being coupled to a critical heat flux ratio. Therefore, the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment know the penalty factor parameter, which is the factor parameter of the predetermined penalty, and the penalty factor If it is possible to calculate a predetermined penalty based on the parameters, penalties other than the bending penalty 68 should be evaluated rationally according to the actual situation, instead of conservative evaluation. An evaluation of the furnace 20 can be made.

In the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment, the evaluation unit 12 uses the calculated predetermined penalty and the uncertainty parameter as thermal It is linked to the critical heat flux ratio of the reactor 20 calculated by flow analysis. Therefore, the nuclear reactor evaluation apparatus 10 according to the embodiment, the nuclear reactor evaluation method according to the embodiment, and the nuclear reactor evaluation program according to the embodiment treat the predetermined penalty in the same manner as the uncertainty parameter, can be coupled to the critical heat flux ratio of

REFERENCE SIGNS LIST 10 reactor evaluation device 12 evaluation section 20 reactor 25 coolant 30 core 32 fuel assembly 34 fuel rod 36 control rod guide thimble 38 instrumentation guide thimble 40 region 51 primary coolant flow rate 52 core bypass flow rate 53 reactor pressure 54 Primary coolant average temperature 55 Reactor power 56 Engineering heat-water channel coefficient 57 Nuclear enthalpy rise heat-water channel coefficient 58 Bending amount 59 Ratio 60, 69 Probability distribution 61 Lower limit 62 Allowable limit 63 Normal value 64, 65 Difference 66 Minimum Critical Heat Flux Ratio 67 Local Quality 68 Bending Penalty 71,72,73,74,75,76,77 Processing 80 Graph 81 First Relationship Curve 82,88 First Threshold 83 Second Threshold 84,85,86,89 Straight line 87 Second relationship curve

Claims (8)

Equipped with an evaluation unit that evaluates the critical heat flux ratio of the reactor,
The evaluation unit
By performing a heat flow analysis in the core of the nuclear reactor, calculating a local quality, which is a parameter representing the quality of thermal equilibrium for each divided region divided into a plurality of regions in the core,
Based on the local quality, determine a local space where the local critical heat flux ratio, which is a local critical heat flux ratio, is the minimum,
Based on the local quality in the local space and the amount of bending of the fuel rods in the reactor, a bending penalty, which is a parameter representing the effect of the bending of the fuel rods on the critical heat flux ratio, is calculated;
A nuclear reactor evaluation apparatus, wherein the calculated bending penalty is combined with the critical heat flux ratio of the nuclear reactor calculated by the thermal hydraulic analysis. The evaluation unit
The bending penalty is calculated using a bending penalty correlation formula having the local quality and the bending amount of the fuel rod as input parameters,
The bending penalty correlation formula is
As the local quality, using a value obtained by subtracting the saturated enthalpy of the liquid phase of the coolant from the enthalpy of the coolant with respect to the latent heat of the coolant,
If the local quality is less than or equal to the first threshold, the bending penalty is calculated as a predetermined positive number, and if the local quality is greater than the first threshold and less than or equal to the second threshold, the predetermined positive number 2 . The nuclear reactor evaluation system according to claim 1 , wherein the calculation is performed using a function that monotonically decreases according to the local quality from , and the calculation is performed with 0 when the local quality is greater than a second threshold value. The amount of bending of the fuel rods is determined by setting the distance between the adjacent fuel rods when the fuel rods having a curve are included and the distance between the adjacent fuel rods when the fuel rods having a curve are not included. If the distance between is D, it is 1-(D'/D),
3. The nuclear reactor evaluation apparatus according to claim 2, wherein the evaluation unit determines the first threshold value of the local quality according to the bending amount. The evaluation unit
4. The atom according to any one of claims 1 to 3, wherein the bending amount of the fuel rod is statistically treated as a random variable, and the bending penalty is calculated using a bending penalty correlation formula. Furnace evaluation device. a local quality calculation step of calculating a local quality, which is a parameter representing the quality of thermal equilibrium for each of a plurality of divided regions in the core, by performing a heat flow analysis in the core of the nuclear reactor;
a local space determination step of determining a local space in which a local critical heat flux ratio, which is a local critical heat flux ratio, is minimized based on the local quality;
Bending penalty calculation for calculating a bending penalty, which is a parameter representing the effect of the bending of the fuel rods on the critical heat flux ratio, based on the local quality in the local space and the amount of bending of the fuel rods in the reactor. a step;
a critical heat flux ratio evaluating step of combining the bending penalty calculated in the bending penalty calculating step with the critical heat flux ratio of the nuclear reactor calculated in the thermal hydraulic analysis;
A nuclear reactor evaluation method comprising: 6. The nuclear reactor evaluation method according to claim 5 , wherein said bending amount of said fuel rod is statistically treated as a random variable, and said bending penalty is calculated using a bending penalty correlation formula. In the evaluation section that evaluates the critical heat flux ratio of the nuclear reactor,
a local quality calculation step of calculating a local quality, which is a parameter representing the quality of thermal equilibrium for each of a plurality of divided regions in the core, by performing a heat flow analysis in the core of the nuclear reactor;
a local space determination step of determining a local space in which a local critical heat flux ratio, which is a local critical heat flux ratio, is minimized based on the local quality;
Bending penalty calculation for calculating a bending penalty, which is a parameter representing the effect of the bending of the fuel rods on the critical heat flux ratio, based on the local quality in the local space and the amount of bending of the fuel rods in the reactor. a step;
a critical heat flux ratio evaluating step of combining the bending penalty calculated in the bending penalty calculating step with the critical heat flux ratio of the nuclear reactor calculated in the thermal hydraulic analysis;
A nuclear reactor evaluation program characterized by executing to the evaluation unit,
8. The nuclear reactor evaluation program according to claim 7 , wherein said bending amount of said fuel rod is statistically treated as a random variable, and said bending penalty is calculated using a bending penalty correlation formula.

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