Cooling performance evaluation method, analysis device, cooling performance evaluation program

Abstract

To evaluate the cooling performance at LOCA time in a smaller calculation amount while maintaining accuracy.SOLUTION: A method includes: a group creation step of grouping fuel rods based on a fuel rod output history; an analysis step of analyzing whether fine pellets are emitted at a loss of coolant accident for each group; and an evaluation step of evaluating a cooling performance by calculating the amount of emitted fine pellets according to the group of fuel rods and the analysis results of whether the fine pellets are emitted.SELECTED DRAWING: Figure 2

Description

Technical Field

[0001] The present disclosure relates to a cooling performance evaluation method, an analysis device, and a cooling performance evaluation program.

Background Art

[0002] As an evaluation of the safety of a nuclear reactor, there is an analysis for evaluating the state when a Loss of Coolant Accident (LOCA) occurs in the nuclear reactor. For example, Patent Document 1 describes a method for evaluating the case where the primary coolant decreases during a small-scale loss of coolant accident using an analysis code that calculates the distribution of non-condensable gas in the primary system.

Prior Art Documents

Patent Documents

[0003]

Patent Document 1

Summary of the Invention

Problems to be Solved by the Invention

[0004] A nuclear reactor needs to be able to maintain its cooling performance even when a loss of coolant accident occurs in the nuclear reactor. In recent years, in tests assuming a loss of coolant accident for fuel rods irradiated to a high burnup, it has been found that the fuel rods are damaged and the pellets are released outside the fuel rods. Thus, even when a loss of coolant accident occurs in a nuclear reactor including fuel rods irradiated to a high burnup and the fuel rods irradiated to a high burnup are damaged, it is necessary to confirm that the cooling performance of the nuclear reactor can be maintained. Here, generally, tens of thousands of fuel rods are loaded in a nuclear reactor, and in order to evaluate the cooling performance at the time of damage, the amount of calculation becomes enormous if an analysis is performed for all the fuel rods. Further, if an analysis is performed assuming that all the fuel rods are fuel rods under the most severe operating conditions, the safety factor becomes too high and the cooling performance of the nuclear reactor cannot be accurately estimated.

[0005] Therefore, an object of the present disclosure is to provide a cooling performance evaluation method, an analysis device, and a cooling performance evaluation program that can evaluate the cooling performance during a loss-of-coolant accident (LOCA) with a small amount of calculation while maintaining accuracy. **Means for Solving the Problems**

[0006] The cooling performance evaluation method of the present disclosure includes a group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing for each group whether the release of fine pellets occurs during a loss-of-coolant accident, and an evaluation step of calculating the amount of fine pellets to be released based on the group of fuel rods and the analysis result of whether the release of fine pellets occurs, and evaluating the cooling performance.

[0007] The analysis device of the present disclosure is an analysis device having an arithmetic unit for evaluating the cooling performance during a loss-of-coolant accident in the reactor core. The arithmetic unit executes a group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing for each group whether the release of fine pellets occurs during a loss-of-coolant accident, and an evaluation step of calculating the amount of fine pellets to be released based on the group of fuel rods and the analysis result of whether the release of fine pellets occurs, and evaluating the cooling performance.

[0008] The cooling performance evaluation program of the present disclosure causes a computer to execute a group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing for each group whether the release of fine pellets occurs during a loss-of-coolant accident, and an evaluation step of calculating the amount of fine pellets to be released based on the group of fuel rods and the analysis result of whether the release of fine pellets occurs, and evaluating the cooling performance. **Advantages of the Invention**

[0009] According to the present disclosure, the cooling performance during a LOCA is evaluated with a small amount of calculation while maintaining accuracy. **Brief Description of the Drawings**

[0010]

Figure 1

Figure 2

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DETAILED DESCRIPTION OF THE INVENTION

[0011] Hereinafter, embodiments of the present invention will be described in detail with reference to the drawings. Note that the present invention is not limited by this embodiment. In addition, the components in the following embodiments include those that can be replaced by those skilled in the art and are easy to replace, or those that are substantially the same. Furthermore, the components described below can be combined as appropriate, and when there are a plurality of embodiments, the embodiments can also be combined.

[0012] The cooling performance evaluation method according to the present embodiment is a method for evaluating the cooling performance of a reactor core when a Loss of Coolant Accident (LOCA) occurs in a nuclear reactor. More specifically, the cooling performance evaluation method analyzes the amount of pellets, which are radioactive substances, released from the fuel rods of a nuclear reactor during a LOCA, and evaluates the cooling performance of the nuclear reactor based on whether the amount of released pellets is an amount that can be cooled. The cooling performance evaluation method performs analysis assuming that a LOCA occurs at the end of the burnup of the fuel rods, that is, immediately before the end of a long-term operation period.

[0013] FIG. 1 is a block diagram schematically showing an analysis device according to the present embodiment. With reference to FIG. 1, the analysis device 1 will be described. The analysis device 1 analyzes the physical damage of fuel rods during LOCA based on the output history of the fuel rods, and evaluates the cooling performance based on the analysis results. The analysis device 1 analyzes the state of the fuel rods during LOCA based on the output history of the fuel rods. The analysis device 1 creates groups of fuel rods based on the output history of the fuel rods. The analysis device 1 simulates the behavior of the fuel rods during LOCA. Specifically, the analysis device 1 performs a LOCA analysis of the reactor core. The analysis device 1 calculates the number of fuel rods from which pellets are released and the amount of pellets released based on the results of the LOCA analysis and the information on the groups of fuel rods. The analysis device 1 evaluates the cooling performance of the nuclear reactor based on the number of fuel rods from which pellets are released and the amount of pellets released.

[0014] The analysis device 1 includes an arithmetic unit 11, a storage unit 12, a display unit 13, and an input unit 14.

[0015] The arithmetic unit 11 includes, for example, an integrated circuit such as a CPU (Central Processing Unit). The arithmetic unit 11 executes analysis processing and the like related to LOCA analysis based on the input information. The storage unit 12 is an arbitrary storage device such as a semiconductor storage device and a magnetic storage device. Various programs for executing various processes, for example, a cooling performance evaluation program, and various data used for the processes are stored in the storage unit 12. Examples of the various data include the arrangement layout information of the fuel rods of the nuclear reactor, the specification information of the fuel rods, the output history information of the fuel rods, the input information (input parameters) D1 input to the LOCA analysis, the output information D2 output as the analysis result, and the like. The display unit 13 is a display device such as a liquid crystal display. The input unit 14 is an input device such as a keyboard and a mouse. Note that the display unit 13 and the input unit 14 may be integrated as an input display device capable of input operations such as a touch panel.

[0016] Next, with reference to FIGS. 2 to 6, a cooling performance evaluation method executed by the analysis device 1 will be described. FIG. 2 is a flowchart showing an example of the processing of the cooling performance evaluation method according to the present embodiment. FIG. 3 is a graph showing an example of the variation in the output of the fuel rod. FIG. 4 is a schematic diagram showing an example of the classification result of the fuel rod group. FIG. 5 is a schematic diagram showing an example of the conditions for burst occurrence of the fuel rod. FIG. 6 is a schematic diagram showing an example of the distribution of the occurrence probability of the burst of the fuel rod.

[0017] The analysis device 1 executes the cooling performance evaluation method shown in FIG. 2 by processing the cooling performance evaluation program stored in the storage unit 12 in the calculation unit 11. The calculation unit 11 selects a plant to be evaluated (step S12). The calculation unit 11 acquires information on the plant to be analyzed, that is, the information on the nuclear reactor, based on the input information D1.

[0018] The calculation unit 11 calculates an allowable value of the cooling capacity (step S14). The calculation unit 11 calculates the cooling performance at the time of LOCA based on the information on the plant to be analyzed. The calculation unit 11 of the present embodiment calculates the number of burst fuel rods that can be cooled at the time of LOCA based on the amount of pellets released from the fuel rods that can be cooled at the time of LOCA. That is, the amount of released pellets that can be cooled by the cooling performance at the time of LOCA is calculated. Further, the calculation unit 11 calculates the number of burst fuel rods that can be cooled based on the amount of pellets. The criterion for determining whether it can be cooled at the time of LOCA is set based on the conditions set for the nuclear reactor.

[0019] The calculation unit 11 acquires information on the output history of the fuel rods to be evaluated (step S16). The calculation unit 11 acquires information on the output history of the fuel rods used in the plant to be analyzed based on the input information D1. Here, since the nuclear fissile uranium in the fuel assembly including the fuel rods decreases with use in the nuclear reactor, for each operation cycle, the fuel assembly with decreased nuclear fissile uranium is removed, a new fuel assembly is loaded, and the loading position in the nuclear reactor is changed so that a predetermined thermal output can be extracted during the period of the new operation cycle. The output for each loading position can be calculated based on the information of the instrumentation tubes arranged in the nuclear reactor and can be calculated by numerical analysis. The calculation unit 11 acquires the information on the output history of the fuel rods calculated based on the information on the output load and the loading position during the operation cycle. The calculation unit 11 does not need to acquire the information on the output history of all the fuel rods, but acquires the information on the fuel rods whose burnup at the end of combustion is equal to or higher than the threshold value, that is, the information on the fuel rods with a high load during use and a high possibility of damage during LOCA. For example, the output history of the fuel rods with a local burnup equal to or higher than the threshold value is extracted. Also, the calculation unit 11 may be a process of extracting the fuel rods that satisfy the conditions, or may be a process of excluding from the analysis target the fuel rods that are determined to have no possibility of damage during LOCA.

[0020] As shown in FIG. 3, the information on the output history shows that when the horizontal axis represents the operation period and the vertical axis represents the output, the output fluctuates in each cycle CY1, CY2, CY3, CY4. The output histories 40, 42, 44 in FIG. 3 are examples, and the output distribution varies for each fuel rod. Also, when comparing the output histories 40, 42, 44 for each cycle CY1, CY2, CY3, CY4, they have the same tendency. The fuel rods to be evaluated have various output histories. For example, there are fuel rods with the largest output in cycle CY1 and fuel rods with the largest output in cycle CY3. Also, FIG. 3 schematically shows the output, and the actual output history fluctuates within each cycle CY1, CY2, CY3, CY4.

[0021] The calculation unit 11 extracts the output history in which the internal pressure of the fuel rod increases (step S18). For the fuel rod for which the calculation unit 11 has acquired the output history information, the internal pressure at the time of LOCA is calculated. The internal pressure of the fuel rod can be calculated by analysis using various analysis codes conventionally used. The calculation unit 11 extracts the output history in which the internal pressure of the fuel rod increases based on the analysis result.

[0022] The calculation unit 11 classifies the fuel rods into groups (step S20). The calculation unit 11 classifies the fuel rods into a plurality of groups using the results of step S16 and step S18. For example, as shown in FIG. 3, fuel rods having the same magnitude relationship of the output compared in terms of the cycle are set as the same group. Also, the groups may be classified by the output of the final loading cycle or the burnup. The number of groups to be classified and the priority order of the classification conditions can be set in advance. The calculation unit 11 classifies the fuel rods to be evaluated into the fuel rods of group 52, the fuel rods of group 54, and the fuel rods of group 56 as shown in FIG. 4 by classifying using the output history of the fuel rods, the burnup (integrated output), the output of the final loading cycle, etc. Note that the number of groups shown in FIG. 4 is three, but it is not limited thereto. As an example, classification can be made into three by the output of the final loading cycle and into five by the integrated output to create 15 groups.

[0023] The calculation unit 11 executes a LOCA analysis of the fuel rod (step S22). The calculation unit 11 performs a sensitivity analysis in which the output before the accident and the internal pressure of the fuel rod are changed using the conditions of the fuel rod and the reactor, and calculates the burst occurrence conditions of the internal pressure of the fuel rod for each output before the accident. For the sensitivity analysis, various analysis codes used in nuclear power analysis can be used. Also, the output before the accident and the internal pressure of the fuel rod are specified by various parameter setting methods such as random sampling and the repeated analysis is executed. The calculation unit 11 calculates the boundary condition 60 of the conditions for burst occurrence and the conditions for avoiding burst in the relationship between the output before the accident and the internal pressure of the fuel rod as shown in FIG. 5 by performing the analysis. Note that the LOCA analysis in step S22 may be calculated in advance by another calculation unit and the result may be acquired.

[0024] The calculation unit 11 evaluates the internal pressure of the fuel rods in the group (step S24). The calculation unit 11 selects a representative fuel rod from the fuel rods in the group. Here, the representative fuel rod is, for example, the fuel rod with the highest internal pressure among the fuel rods in the group. That is, it is the fuel rod determined to be the most likely to be damaged among the fuel rods in the group. For example, in FIG. 3, the output history 40 is the output history of the representative fuel rod. The calculation unit 11 calculates the pre-accident internal pressure of the fuel rod based on the output history of the representative fuel rod. Based on the calculation result of the internal pressure of the representative fuel rod and the uncertainty, the internal pressure distribution of the fuel rods in the group is calculated. In the present embodiment, the uncertainty of the internal pressure evaluation of the fuel rod is enclosed within 25% of the most accurate value, and the uncertainty of 25% is a normal distribution with a 95% confidence interval, that is, 2σ. As a result, as shown in FIG. 6, the probability distribution of the fuel rod internal pressure can be calculated. The distribution 70 shown in FIG. 6 has the internal pressure of the representative fuel rod as the maximum value of the probability distribution, and the position of the value 74 that is 25% greater than the internal pressure of the representative fuel rod is the distribution with 2σ of the normal distribution. Note that setting the distribution of the fuel rods in the present embodiment as a normal distribution and setting the uncertainty of 25% at the position of 2σ is an example, and the distribution setting is not limited to this. The distribution of the fuel rods may be evaluated based on measured values calculated in advance (the measured values of fuel rods with a shorter operation period than the present embodiment may also be used), and may be a distribution other than the normal distribution. Also, the position of the 25% uncertainty may be set at the position of σ or 3σ instead of the position of 2σ.

[0025] The calculation unit 11 calculates the number of burst fuel rods in the group (step S26). The calculation unit 11 determines the condition 72 for burst occurrence among the distributions 70 in FIG. 6 based on the burst boundary condition 60 shown in FIG. 5, and the fuel rods in the region part higher than the condition 72 are regarded as burst fuel rods. The condition 72 is such that the lower the pre-accident output of the representative fuel rod in the group, the higher the internal pressure of the fuel rod for the burst occurrence condition. The calculation unit 11 calculates the number of burst fuel rods in the group by integrating the ratio of the region part higher than the condition 72 and the number of fuel rods in the group.

[0026] The calculation unit 11 determines whether the analysis of all groups has been completed (step S28). When the calculation unit 11 determines that the analysis of all groups has not been completed (No in step S28), it returns to step S22 and performs the analysis of the unanalyzed groups.

[0027] When the calculation unit 11 determines that the analysis of all groups has been completed (Yes in step S28), it calculates the total number of burst rods (step S30). The calculation unit 11 calculates the total number of burst rods of the fuel rods in each group calculated in the processes from step S22 to step S26.

[0028] The calculation unit 11 determines whether it can be cooled based on the calculation result (step S32). The calculation unit 11 determines whether the total number of burst rods calculated in step S30 exceeds the allowable value of the cooling performance calculated in step S14. When the total number of burst rods calculated in step S30 is less than the number of burst rods of the allowable value of the cooling performance calculated in step S14, the calculation unit 11 determines that it can be cooled. When the total number of burst rods calculated in step S30 is greater than or equal to the number of burst rods of the allowable value of the cooling performance calculated in step S14, the calculation unit 11 determines that it cannot be cooled.

[0029] When the calculation unit 11 determines that it cannot be cooled (No in step S32), that is, when it is determined that the number of fuel rods exceeding the allowable value of the cooling performance will burst, it returns to step S20 and re-executes the grouping of the fuel rods. The calculation unit 11 increases the number of groups to be classified compared to the number of groups in the process where it is determined that it cannot be cooled. That is, the calculation unit 11 analyzes the fuel rods in more subdivided groups. When the calculation unit 11 determines that it can be cooled (Yes in step S32), this process ends.

[0030] The analysis device 1 can evaluate the cooling performance of the pellets released into the reactor during a LOCA of the nuclear reactor with high accuracy by dividing the fuel rods into a plurality of groups based on the output history of the fuel rods and evaluating the burst of the fuel rods for each group. Further, the analysis device 1 can suppress an increase in the calculation load by evaluating the burst of the fuel rods for each group. Further, the analysis device 1 can evaluate that the reactor can be safely operated even for an output history that is determined not to satisfy the cooling performance when evaluated based on the most severe fuel rods by evaluating the burst of the fuel rods for each group.

[0031] Further, the analysis device 1 can evaluate the cooling performance under severe conditions and improve the reliability as a safety evaluation by evaluating the most severe fuel rod in the group as a representative fuel rod, that is, by using the fuel rod that is most likely to burst in the group as the representative fuel rod.

[0032] Further, when it is determined that the cooling performance is not satisfied, by increasing the number of groups and performing the analysis, although the amount of calculation increases, the analysis can be performed in a state closer to the actual state.

[0033] In the above embodiment, the cooling performance was evaluated based on the number of burst fuel rods. However, instead of the number of burst fuel rods, the cooling performance may be evaluated based on the amount of pellets (fuel) released from the fuel rods. The analysis device 1 can determine whether the operating conditions of the nuclear reactor satisfy the conditions by evaluating whether the fuel released from the fuel rods during a LOCA can be cooled.

[0034] As described above, the present disclosure can be understood as follows, for example. (1) A group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing for each group whether the release of fine pellets occurs during a coolant loss accident, and calculating the amount of the fine pellets to be released based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs, and an evaluation step of evaluating the cooling performance. Thereby, the cooling performance for the pellets released into the reactor during a LOCA of the nuclear reactor can be evaluated with high accuracy. Also, by evaluating the burst of the fuel rods for each group, an increase in the calculation load can be suppressed.

[0035] (2) It has a number calculation step of calculating the number of bursts of the fuel rods that can be cooled in the nuclear reactor to be analyzed, and the evaluation step calculates the number of the fuel rods that burst based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs, and evaluates the cooling performance by comparing with the number of bursts of the fuel rods that can be cooled calculated in the number calculation step. The cooling performance evaluation method according to (1). Thereby, the cooling performance for the pellets released into the reactor during a LOCA of the nuclear reactor can be evaluated with high accuracy.

[0036] (3) The group creation step classifies the groups based on the internal pressure evaluation of the fuel rods. The cooling performance evaluation method according to (1) or (2). Thereby, the groups can be classified based on the state of the fuel rods.

[0037] (4) The analysis step performs the analysis with the output history that makes the fuel rod internal pressure the highest among the fuel rods included in the group. The cooling performance evaluation method according to (3). Thereby, the cooling performance can be evaluated under severe conditions, and the reliability can be increased as an evaluation of safety.

[0038] (5) In the evaluation step, if it is determined that the cooling performance does not meet the conditions, return to the group creation step, and the group creation step increases the number of groups to be created and performs grouping according to any one of (1) to (4) in the cooling performance evaluation method described above. Thereby, the cooling performance of the pellets released into the reactor during a LOCA of the nuclear reactor can be evaluated with high accuracy. Also, the workload of the operator can be reduced.

[0039] (6) An analysis device having a calculation unit for evaluating the cooling performance during a coolant loss accident of the reactor core, wherein the calculation unit includes a group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing whether the release of fine pellets occurs during a coolant loss accident for each group, and an evaluation step of calculating the amount of the fine pellets to be released and evaluating the cooling performance based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs. Thereby, the cooling performance of the pellets released into the reactor during a LOCA of the nuclear reactor can be evaluated with high accuracy. Also, by evaluating the burst of the fuel rods for each group, an increase in the calculation load can be suppressed.

[0040] (7) A cooling performance evaluation program that causes a computer to execute a group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing whether the release of fine pellets occurs during a coolant loss accident for each group, and an evaluation step of calculating the amount of the fine pellets to be released and evaluating the cooling performance based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs. Thereby, the cooling performance of the pellets released into the reactor during a LOCA of the nuclear reactor can be evaluated with high accuracy. Also, by evaluating the burst of the fuel rods for each group, an increase in the calculation load can be suppressed.

Explanation of symbols

[0041] 1 Analysis device 11 Calculation unit 12 Storage unit 13 Display unit 14 Input unit D1 Input information D2 Output information

Claims

1. A group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing, for each group, whether the release of fine pellets occurs during a coolant loss accident, and an evaluation step of calculating the amount of the fine pellets to be released based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs, and evaluating the cooling performance. A cooling performance evaluation method comprising the above steps.

2. A number calculation step of calculating the number of bursts of the fuel rods that can be cooled in the reactor to be analyzed, wherein the evaluation step calculates the number of the fuel rods that burst based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs, and evaluates the cooling performance by comparing with the number of bursts of the fuel rods that can be cooled calculated in the number calculation step. The cooling performance evaluation method according to Claim 1.

3. The group creation step classifies the groups based on the internal pressure evaluation of the fuel rods. The cooling performance evaluation method according to Claim 1.

4. In the analysis step, the analysis is performed with the output history that results in the highest internal pressure of the fuel rods among the fuel rods included in the group. The cooling performance evaluation method according to Claim 3.

5. If it is determined in the evaluation step that the cooling performance does not meet the conditions, the process returns to the group creation step. The group creation step increases the number of groups to be created and performs grouping. The cooling performance evaluation method according to any one of Claims 1 to 4.

6. An analysis device having a calculation unit for evaluating the cooling performance during a coolant loss accident in the reactor core, wherein the calculation unit performs a group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing, for each group, whether the release of fine pellets occurs during a coolant loss accident, and an evaluation step of calculating the amount of the fine pellets to be released based on the group of the fuel rods and the analysis result of whether the release of the fine pellets occurs, and evaluating the cooling performance. An analysis device that executes the above steps.

7. A group creation step of grouping the fuel rods based on the output history of the fuel rods, an analysis step of analyzing, for each group, whether the release of fine pellets occurs during a coolant loss accident, A cooling performance evaluation program that causes a computer to execute an evaluation step of calculating the amount of the fine pellets to be discharged based on the group of the fuel rods and the analysis result of whether the discharge of the fine pellets occurs, and evaluating the cooling performance.

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