Reliability improvement method of nuclear power plant

Abstract

To provide a reliability improvement method of a nuclear power plant capable of improving the reliability of a nuclear power plant by using corrosion potential as continuous monitoring indicators as a performance indicator of reliability-oriented maintenance processes.SOLUTION: A disclosed reliability improvement method of a nuclear power plant P100 includes: a selection step S1 of selecting a device or part with a high risk of stress corrosion cracking as a device or part that is of high importance for maintenance; a monitoring step S2 of monitoring the corrosion potential of the selected device or part that is of high importance for maintenance as a monitoring indicator and constantly monitoring the crack growth rate at the monitored area by using the corrosion potential based on water quality parameters and plant data during the driving cycle; and a preventive maintenance execution step S3 of determining the timing of the next preventive maintenance inspection by using the ratio of the constantly monitored crack growth rate and the estimated crack growth rate in the device or part with the high risk of stress corrosion cracking when continuous monitoring is not performed.SELECTED DRAWING: Figure 1

Description

[Technical field]

[0001] The present invention relates to a method for improving the reliability of a nuclear power plant. [Background technology]

[0002] Structural components such as equipment and piping in nuclear power plants (hereafter sometimes referred to as nuclear plants) are made of structural materials such as stainless steel and nickel-based alloys. These structural materials show susceptibility (including information on the initiation and propagation rate of cracks and cracks) to stress corrosion cracking (hereafter referred to as SCC) when three factors, namely material, stress, and corrosive environment, meet certain conditions, and SCC will initiate and propagate.

[0003] Of the three factors above, many of the conditions for materials and stress are determined at the time of design and construction. Some of them change over time due to radiation exposure and aging, but it is difficult to confirm these material and stress characteristics during operation. On the other hand, the corrosive environment is determined by the operation of the plant. In a boiling water reactor (BWR), structural materials are exposed to the cooling water (reactor water) in the pressure vessel and recirculation system piping. The reactor water contains oxygen and hydrogen peroxide, which cause corrosion and are generated by radiolysis of the reactor water in the pressure vessel, as well as various ions as impurities that accelerate corrosion. The reactor water containing impurities as well as oxygen and hydrogen peroxide forms a corrosive environment for structural materials. Some of these parameters can be measured during operation by instruments installed in the sampling system.

[0004] Furthermore, with regard to SCC, the corrosion potential (ECP) of structural materials is widely used as an index of the strength of the corrosive environment. When no technology (corrosive environment mitigation technology) is applied to mitigate the corrosive environment and control the initiation and progression of SCC (normal water quality), it is known that the ECP of structural materials exposed to BWR reactor water usually shows a high value of 0 mV vs. SHE or higher. On the other hand, under BWR reactor water conditions, it is known that the initiation of SCC is suppressed when the ECP is lower than a value of about -300 to -200 mV vs. SHE. It is also known that a decrease in ECP also reduces the crack growth rate. Therefore, hydrogen injection (HWC) or noble metal injection (NMCA) are widely used as corrosive environment mitigation technology both in Japan and overseas.

[0005] In addition, when applying corrosive environment mitigation technology, it is necessary to measure the ECP of structural materials with high accuracy to confirm that the SCC suppression effect is obtained. For this reason, ECP sensors are installed in pressure vessels, in piping connected to the pressure vessels, or in piping drawn from them to measure the ECP of structural materials. In many cases, they are installed only when the corrosive environment mitigation technology is introduced and removed after the effect is confirmed, or they are installed in piping outside the pressure vessel for measurement.

[0006] Although ECP is the main factor influencing SCC, it is also strongly influenced by impurity ions, especially chloride and sulfate ions, contained in the reactor water. In other words, even if the ECP is the same, the higher the concentration of harmful impurities in the reactor water, the more likely SCC will occur. Therefore, it can be said that the susceptibility to SCC occurrence and progression depends not only on ECP but also on the impurity ion concentration. Electrical conductivity is determined by the concentration of impurities contained in the reactor water and the mobility of the ions. For this reason, in BWRs, operation has been carried out to keep the purity of the reactor water high by lowering the impurity ion concentration (pure water management). On the other hand, if the ECP is sufficiently low, for example, at about -400mV vs.SHE, the effect of increased electrical conductivity on SCC has a margin, and it is also known that even if the electrical conductivity due to impurity ions increases from the normal 10μS / m to about 20μS / m, the susceptibility to SCC is small. Therefore, pure water management is important for SCC management in BWRs with normal water quality.

[0007] Under such circumstances, a technique is known for monitoring SCC by ECP and evaluating crack length taking into account the time to SCC onset (see, for example, Patent Document 1). Patent Document 1 describes a method for measuring the ECP of a structural material in contact with cooling water containing impurity ions and the concentration of the impurity ions contained in the cooling water, determining the time to SCC onset in the structural material based on this information, and reducing at least one of the ECP and the concentration of the impurity ions when this onset time is shorter than a set time. This technique also describes that the crack growth management provides a margin for time by calculating the time to SCC onset from the plant state including the ECP and the impurity ion concentration.

[0008] Also, a system for predicting the crack growth life due to SCC is known (see, for example, Patent Document 2). The life prediction described in Patent Document 2 is performed using measured ECP and crack growth characteristic data of the structural material.

[0009] Other methods for estimating the remaining life of structural components of nuclear power plants are also known (see, for example, Patent Document 3). The remaining life estimation method described in Patent Document 3 monitors or estimates the dissolved oxygen concentration (O2), hydrogen peroxide concentration (H2O2), and electrical conductivity (μ) in the reactor water of a portion of a structure in a nuclear reactor for which the remaining life is to be estimated. Then, using these values, the method obtains the growth rate of a crack caused by stress corrosion for a crack assumed to exist in the portion for which the remaining life is to be estimated, and calculates the time until the crack reaches a predetermined limit value from the growth rate of the crack, which is then used as the remaining life. [Prior art documents] [Patent documents]

[0010] [Patent Document 1] Patent No. 5483385 [Patent Document 2] JP 2009-30980 A [Patent Document 3] Japanese Patent Application Publication No. 6-34786 Summary of the Invention [Problem to be solved by the invention]

[0011] When managing a nuclear power plant as described in Patent Documents 1 to 3, the occurrence and growth of SCC cracks are evaluated, predicted, estimated, etc., and then various countermeasures are applied as necessary. For example, countermeasures including stress improvement, including corrosive environment mitigation technology for SCC, material replacement, and SCC crack inspection technology are applied as measures to suppress the occurrence and growth of SCC, repair discovered SCC cracks, and post-repair measures. Conventionally, these were mainly applied individually. In addition, when operating while allowing cracks, it was necessary to evaluate future crack growth and then continuously check the growth state of the cracks by inspection at each regular inspection after the cracks were discovered.

[0012] Meanwhile, the introduction of reliability-centered maintenance (RCM) is being promoted worldwide in nuclear plants. A representative example of RCM for nuclear plants is AP913 (equipment reliability process) by the Institute of Nuclear Power (INPO). In RCM, in order to improve equipment reliability (ER), an optimal maintenance program is formulated by considering the selection of important structures, systems, and equipment (systems, structures, and components; hereafter, these may be collectively referred to as "SSC"), the selection of optimal maintenance methods, and the selection of implementation timing. It is necessary to aim to improve the reliability of the entire nuclear plant by introducing such RCM. However, currently, there is no RCM process for the management of SCC, which has a significant impact on the ER of nuclear plants. Therefore, as mentioned above, SCC countermeasure technologies are limited to being applied individually.

[0013] By linking SCC management with RCM, it is believed that it will be possible to link individual measures and SCC management procedures and build a process for long-term, continuous improvement of ER throughout the entire life cycle of a nuclear plant. This is expected to further improve the ER of nuclear plants. In the RCM process, in order to indicate the ER state, specific performance indicators must be constantly monitored and clearly indicated (constant monitoring refers to continuous monitoring that grasps the temporal changes of the monitored performance indicators over a certain period, including temporary interruptions of monitoring due to maintenance of the monitoring means or failures. It does not mean monitoring without interruption for even one second during a set period.); however, since there was no RCM process for SCC, constant monitoring indicators had not been determined.

[0014] The present invention has been made in view of the above circumstances, and an object of the present invention is to provide a method for improving the reliability of a nuclear power plant, which can use the corrosion potential (ECP) as a performance index of a reliability-based maintenance (RCM) process and can improve the ER of the nuclear power plant by using the ECP as a continuous monitoring index. [Means for solving the problem]

[0015] In order to solve the above problems, the present invention embodies an ER process for SCC, which is characterized by using ECP as a performance index for constantly monitoring and indicating ER for SCC, and continuously evaluating the crack growth rate of SCC over the medium to long term by inputting the ECP values ​​monitored along with the materials used for each SSC and the stress state in welds, etc.

[0016] In particular, in the conventional technology and the HWC and NMCA described in the above-mentioned Patent Document 1, the purpose of measuring ECP is to lower the ECP and ensure the effect of lowering the crack growth rate caused by the lowering of the ECP. However, in the present invention, the purpose of constant monitoring is not limited to the lowering of the ECP during plant operation, but is to simply monitor the ECP value including the normal water quality at all times and evaluate the crack growth rate to keep the crack growth rate of SCC in a controlled state. In other words, constant monitoring of ECP brings the cracks into a controlled state, thereby improving the ER. On the other hand, although ECP can be constantly monitored during operation as an influencing factor of SCC, it is generally not measured when environmental mitigation technology is not applied. Therefore, for ECP when environmental mitigation technology is not applied, a value set based on the measurement results in plants that apply environmental mitigation technology and the evaluation results of analysis models is used. Therefore, when ECP is not measured, the value of the crack growth rate is simply a conservative estimate, and it cannot be said that it is in a controlled state.

[0017] Specifically, the present invention is a method for improving the reliability of a nuclear plant, comprising: a selection step of selecting equipment or parts having a high risk of stress corrosion cracking as equipment or parts having high maintenance importance; a monitoring step of monitoring the corrosion potential of the selected equipment or parts having high maintenance importance as a monitoring index and continuously monitoring the crack growth rate at the monitored parts during an operation cycle using the corrosion potential based on water quality parameters and plant data; and a preventive maintenance execution step of determining the timing of the next inspection for preventive maintenance using the ratio between the continuously monitored crack growth rate and a crack growth rate estimated in the equipment or parts having a high risk of stress corrosion cracking when continuous monitoring is not performed. Effect of the Invention

[0018] According to the present invention, it is possible to provide a method for improving the reliability of a nuclear power plant, which can use ECP as a performance index of an RCM process as a constantly monitored index, thereby improving the reliability of the nuclear power plant. Problems, configurations and effects other than those described above will become apparent from the following description of the embodiments. [Brief description of the drawings]

[0019] [Figure 1] 1 is a block diagram showing a functional flow of a method for improving reliability of a nuclear power plant according to an embodiment of the present invention; [Diagram 2] FIG. 2 is a block diagram showing a functional flow linking a method for improving reliability of a nuclear power plant and AP-913 according to an embodiment of the present invention. [Diagram 3] 1 is an overall system configuration diagram showing an example of a nuclear power plant to which a method for improving reliability of a nuclear power plant according to an embodiment of the present invention is applied. [Figure 4] FIG. 2 is an explanatory diagram showing an example of an effect (a graph of ER improvement by constant monitoring of crack growth rate) obtained by the method for improving reliability of a nuclear power plant according to one embodiment of the present invention over time. [Diagram 5] 1 is a diagram showing an example of values ​​of monitoring indexes over time in a nuclear power plant to which a method for improving nuclear power plant reliability according to an embodiment of the present invention is applied. [Figure 6] FIG. 2 is an overall system configuration diagram showing another example of a nuclear power plant to which a method for improving reliability of a nuclear power plant according to an embodiment of the present invention is applied. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0020] Hereinafter, a method for improving reliability of a nuclear power plant according to an embodiment of the present invention (hereinafter, may be referred to as "this improvement method") will be described in detail with reference to the drawings as appropriate. In the following description, the same components are denoted by the same reference numerals, and duplicated descriptions may be omitted.

[0021] ≪This improvement method≫ Figure 1 is a block diagram showing the functional flow in this improvement method. A block diagram showing the functional flow in RCM linking AP-913 and SCC is shown in Figure 2, but since the relationships between the steps are complicated, the main flow is extracted and shown in Figure 1. Figure 2 is a block diagram showing the functional flow linking this improvement method and AP-913.

[0022] As shown in Figure 1, this improvement method includes a step (selection step) S1 of classifying the importance of SSCs from the perspective of SCC maintenance, a step (monitoring step) S2 of monitoring performance during operating cycles, a step (preventive maintenance execution step) S3 of formulating an inspection and maintenance plan and deciding the timing of the next inspection (preventive maintenance execution step) S4 of replacing, repairing, and carrying out preventive maintenance on SSCs in which cracks are found (improvement step) S4, and a step (life cycle management step) S5 of implementing a medium- to long-term maintenance plan for the next cycle and beyond.

[0023] In the selection step S1, equipment or parts with a high risk of SCC are selected from the perspective of importance of maintenance, based on data from the time of design and construction of the plant, past inspection data, and past maintenance history and SCC information from reactors of the same design. In other words, even if the equipment is important from the perspective of safety, if it is made of materials with low SCC susceptibility in design, and stress improvements such as peening and water-cooled welding have been made during construction, and there is no tensile residual stress, the risk of SCC occurrence and progression is low, so the importance of maintenance is reduced. In addition, parts that have been subjected to polishing or peening as preventive maintenance after construction also have a reduced SCC risk.

[0024] In the monitoring step S2, the ECP of the selected equipment or parts with high conservation importance is monitored as a constant monitoring index, and the crack growth rate is estimated from the result. The crack growth rate is used as a performance index to constantly monitor during the operation cycle. Plant data such as materials and stress, and water quality parameters such as impurity ions and electrical conductivity are also used to estimate the crack growth rate. However, it should be noted that even if the impurity ions and electrical conductivity are measured periodically or constantly monitored, they are average data of the reactor water measured at the sampling line. In addition, the concentration of oxidizers such as oxygen and hydrogen peroxide gives information on the reactor water near the inlet of the sampling line, but since most of the hydrogen peroxide is decomposed into oxygen by thermal decomposition and contact decomposition on the pipe surface while flowing through the sampling line, the concentration inside the reactor cannot be known. Therefore, it is preferable to use the oxidizer concentration as a supplementary index after obtaining a correlation with the ECP of the target part.

[0025] In the preventive maintenance execution step S3, the target equipment or target area is inspected according to a plan to evaluate the presence or absence of cracks and the amount of crack growth. In this step, based on the inspection results and the crack growth rate evaluated (estimated) in the monitoring step S2, it is decided whether or not to continue operation with the crack remaining and when to schedule the next inspection.

[0026] Specifically, in the preventive maintenance execution step S3 in this embodiment, the next inspection timing for preventive maintenance is determined using the ratio between the constantly monitored crack growth rate and the crack growth rate estimated in the equipment or part with high SCC risk when the constant monitoring is not performed. This improvement method provides maintainability that can rationally reduce the inspection frequency by using this ratio, and it becomes possible to reduce the inspection frequency while improving the ER of the plant. This will be described in detail later.

[0027] Furthermore, in the improvement step S4, if the inspection results in the preventive maintenance execution step S3 indicate that the amount of crack growth may exceed an allowable value, the equipment (e.g., piping, etc.) or part containing the part with the crack is replaced, or the cracked part is repaired, and preventive maintenance such as stress improvement is implemented. This suppresses the occurrence and growth of cracks. This also reduces the risk determined in the selection step S1, and improves the reliability of the equipment or part.

[0028] In the life cycle management step S5, a plan is formulated to repeat the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, and the improvement step S4 in the medium to long term, and this plan is executed. In other words, in the single flow of the steps S1 to S4, the procedure is a procedure that determines how to respond to cracks discovered at that time, but by responding within a medium to long term plan in the step S5, it is possible to improve the ER by optimizing the overall maintenance, including responses several cycles ahead, rather than responding during the regular inspection in which the cracks were discovered.

[0029] In this improvement method, measures against SCC are taken in the preventive maintenance execution step S3 and the improvement step S4. Therefore, in this improvement method, by repeating such an ER process for the nuclear plant in the medium to long term, the risk of the equipment or parts with high maintenance importance (high risk) selected in the selection step S1 decreases over time, and the risk ranking of each part of the SSC changes. In other words, in this improvement method, by repeating the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, and the improvement step S4, the equipment or parts with high risk are always monitored, and measures are continuously taken. Therefore, in this improvement method, the risk of the entire SSC of the nuclear plant decreases to a reasonably implementable level (ALARP). In this improvement method, the crack growth rate is evaluated using the ECP through constant monitoring of the ECP, and the plant can be operated in a state where the crack growth rate is controlled, that is, in a state where the ER against SCC is improved. As a result, it is possible to know the presence or absence of a crack obtained by the inspection and the length of the crack, and what kind of change in ECP or electrical conductivity during operation caused it. If the crack length as the integral of the crack growth rate is shorter than predicted by the evaluation model or does not occur, safety can be reasonably corrected by actual measurement to correct the conservatism of the evaluation model. For example, if the crack length calculated by the conservatism of the evaluation formula for the crack growth rate becomes larger than the crack length obtained by inspection, the ECP is a value monitored constantly, so it is considered to be due to the conservatism of the material or stress, and is not due to uncertainty that occurred during operation, so the crack growth rate can be corrected.

[0030] In addition, in FIG. 2, there is a corrective step S6 that is not shown in FIG. 1. This step corresponds to a step in which a prompt note is made when a parameter that influences the crack growth rate as a performance index deviates from a control value. For example, this corresponds to a case in which the electrical conductivity exceeds the applicable range of the evaluation formula for the crack growth rate. In that case, it is necessary to quickly identify the cause in accordance with the provisions of the water quality management guidelines and take measures against, for example, the intrusion of ions. In FIG. 1, this corrective step S6 is treated as being appropriately implemented and is included in the monitoring step S2.

[0031] Examples of reactors to which the present improvement method can be applied include, but are not limited to, boiling water reactors (BWRs), advanced boiling water reactors (ABWRs), pressurized water reactors, natural uranium graphite-moderated gas-cooled reactors, advanced gas-cooled reactors, high-temperature gas reactors, heavy water-moderated reactors, molten salt reactors, and fast breeder reactors. Hereinafter, a configuration example in which the present improvement method is applied to a nuclear power plant (nuclear power plant) using a BWR will be described. Then, specific contents of the present improvement method will be described based on the configuration example. Figure 3 is an overall system configuration diagram showing an example of a nuclear power plant to which the present improvement method is applied.

[0032] <An example of a configuration in which this improvement method is applied to a nuclear power plant> As shown in FIG. 3, a nuclear power plant P100 using a BWR includes a BWR reactor P1, a turbine P3, a condenser P4, and a reactor cleanup system and a feedwater system, which will be described later. The reactor P1 is installed in a reactor containment vessel P11. The reactor P1 has a pressure vessel P12 that houses a reactor core P13. A cylindrical shroud P15 installed in the pressure vessel P12 surrounds the reactor core P13. A plurality of fuel assemblies (not shown) are loaded in the reactor core P13. Each fuel assembly includes a plurality of fuel rods filled with a plurality of fuel pellets made of nuclear fuel material. An annular downcomer P17 is formed between the inner surface of the pressure vessel P12 and the outer surface of the shroud P15. In addition, a plurality of jet pumps P21 are installed at the bottom of the pressure vessel P12. The outlet of the jet pump P21 is disposed at the bottom of the downcomer P17 and communicates with the lower space (lower plenum (not shown)) of the pressure vessel P12. The bottom of the lower plenum is called the furnace bottom. The structure that supports the shroud P15 is the shroud support P41.

[0033] The feedwater system is composed of a condensate pump P5, a condensate purification device P6, a feedwater pump P7, a low-pressure feedwater heater P8, and a high-pressure feedwater heater P9, which are installed in this order on a feedwater piping P10 connecting the condenser P4 and the pressure vessel P12, from the condenser P4 to the pressure vessel P12.

[0034] The reactor cleanup system is configured by installing, in this order, a cleanup system isolation valve P23, a cleanup system pump P24, a regenerative heat exchanger P25, a non-regenerative heat exchanger P26, and a reactor water cleanup device P27 in cleanup system piping P20 that connects the pressure vessel P12 and the feedwater piping P10. The cleanup system piping P20 is connected to the feedwater piping P10 between the high-pressure feedwater heater P9 and the pressure vessel P12.

[0035] Reactor water present in the downcomer P17 in the pressure vessel P12 is sucked into the jet pump P21 and guided to the lower plenum below the reactor core P13. The reactor water is supplied from the lower plenum to the reactor core P13 and heated by heat generated by nuclear fission of the nuclear fuel material contained in the fuel rods of the fuel assemblies. A part of the heated reactor water becomes steam. This steam is guided from the pressure vessel P12 through the main steam pipe P2 to the turbine P3, rotating the turbine P3. A generator (not shown) connected to the turbine P3 is rotated, generating electricity. The steam discharged from the turbine P3 is condensed in the condenser P4 to become water.

[0036] This water is supplied as feedwater through the feedwater piping P10 into the pressure vessel P12. The feedwater flowing through the feedwater piping P10 is pressurized by the condensate pump P5, impurities are removed by the condensate purification device P6, the pressure is further increased by the feedwater pump P7, and the low-pressure feedwater heater P8 and the high-pressure feedwater heater P9 are heated. In addition, an extraction piping P14 is connected between the turbine P3 and the high-pressure feedwater heater P9. Extraction steam extracted from the main steam piping P2 and the turbine P3 by the extraction piping P14 is supplied to the high-pressure feedwater heater P9 by the extraction piping P14, and becomes a heating source for the feedwater. The drain of the extraction steam from the high-pressure feedwater heater P9 is returned to the low-pressure feedwater heater P8. The drain of the extraction steam from the low-pressure feedwater heater P8 is returned to the condensate pump P5.

[0037] The reactor water in the pressure vessel P12 contains metal corrosion products contained in the feed water and products generated by corrosion of structural materials in the pressure vessel P12. Therefore, a certain percentage of the reactor water is purified by the reactor water purification system. The reactor water in the pressure vessel P12 is supplied to the regenerative heat exchanger P25 and the non-regenerative heat exchanger P26 through the purification system piping P20 branched off from the recirculation system piping P30 by the drive of the purification system pump P24, and is cooled to about 50°C by these heat exchangers. The cooled reactor water passes through the reactor water purification device P27, where the metal corrosion products contained in the reactor water are removed, and after being heated by the regenerative heat exchanger P25, it is merged with the feed water flowing in the feed water piping P10 and supplied to the pressure vessel P12.

[0038] When the operation of the reactor P1 is stopped, all the control rods (not shown) are inserted into the core P13. By inserting all the control rods, the nuclear fission reaction of the nuclear fuel material is stopped, and the operation of the reactor P1 is stopped. The heat remaining in the equipment in the core P13 and the pressure vessel P12 is removed by the evaporation of the reactor water, but when the temperature drops to a certain degree, the efficiency of heat removal by the evaporation of the reactor water decreases. Therefore, when the reactor water temperature drops to about 150°C, the equipment in the core P13 and the pressure vessel P12 is cooled using a residual heat removal system (not shown).

[0039] As shown in Figure 3, reactor water from the bottom of the downcomer P17 flows through the cleanup system piping P20 and the recirculation system piping P30. In the nuclear power plant P100, ECP is constantly measured in order to use the SCC crack growth rate as a constant monitoring index for the ER process. For this reason, in the nuclear power plant P100, a corrosion potential sensor P35a is installed in the cleanup system piping P20, and a corrosion potential sensor P35b is installed in the recirculation system piping P30.

[0040] Specifically, in the nuclear power plant P100, a bottom drain line P34 is provided between the bottom of the pressure vessel P12 and the purification system piping P20, which draws out reactor water from the bottom of the pressure vessel P12 and passes the water through the purification system piping P20. In the nuclear power plant P100, a branch line P34a is provided from the bottom drain line P34. A flange P36 is provided on the branch line P34a, and an electrochemical corrosion potential sensor P35a is mounted on the flange P36. In the nuclear power plant P100, the electrochemical corrosion potential sensor P35a measures the ECP of the water quality at the bottom of the pressure vessel P12. The branch line P34a is connected to a sampling line (not shown) for the reactor water, etc. Further, a flange P36 is installed in the recirculation system piping P30, and a corrosion potential sensor P35b is mounted on the flange P36 to measure the ECP of the reactor water flowing through the piping.

[0041] Furthermore, the nuclear power plant P100 may have a corrosion potential sensor P35c installed in the lower plenum of the pressure vessel P12, and a corrosion potential sensor P35d installed in the core P13. The corrosion potential sensor P35c and the corrosion potential sensor P35d may be installed in a neutron instrumentation tube P38. The corrosion potential sensor P35c is installed at the position of the lower plenum inside the neutron instrumentation tube P38. The corrosion potential sensor P35d is installed at the position of the core P13 inside the neutron instrumentation tube P38. Then, holes are drilled in the neutron instrumentation tube P38 so that reactor water in a target region flows into these corrosion potential sensors P35c and P35d, thereby measuring the ECP.

[0042] The structural materials of the equipment and piping of the reactor P1 are made of austenitic stainless steel, for example, SUS316L. In addition to SUS316L, the structural materials of the reactor P1 include SUS304, which is an austenitic stainless steel, SUS304L, which is SUS304 with a reduced carbon content, and SUS316NG (NG: for nuclear power use), which is SUS316L with added nitrogen to increase its strength. It is possible to use these materials to construct various electrodes. Furthermore, for example, the shroud P15 is made of SUS304L, SUS316L, etc. The purification system piping P20 and the recirculation system piping P30 are made of SUS304, SUS316NG, etc. Carbon steel is also used for the purification system piping P20. In addition, the reactor P1 can also use high-nickel alloys as structural materials. The bottom of the reactor requires high strength, so 600-series nickel-based alloys (600 alloys) are used. Similarly, 600 series weld metals (Alloy 182, Alloy 82, Alloy 132, etc.) are used as weld metals in the welds and overlays at the bottom of the furnace.

[0043] <<Specific contents of this improvement method (ER process for SCC)>> Next, returning to Fig. 1, the specific contents of this improvement method (ER process for SCC) will be described. The specific contents of this improvement method will be described by taking an example of application to a nuclear power plant P100 using a BWR. This improvement method can be similarly applied to pressurized water reactors and other nuclear reactors by simply changing the ECP constant monitoring parts according to the design conditions of the reactor. For example, as the constant monitoring parts of the ECP in a pressurized water reactor, when a sensor is installed inside the reactor, a corrosion potential sensor can be inserted from the top of the reactor to the top of the core using a thermocouple tube, or inserted into the core. Also, a corrosion potential sensor can be installed in the sampling line of the steam generator installed in the pressure vessel of the pressurized water reactor.

[0044] The ER process for SCC using this improvement method starts with selection step S1. In selection step S1, equipment or parts with a high risk of SCC are selected. In this selection step S1, the SSC of the nuclear plant P100 pre-selects equipment or parts with a high risk of SCC based on criteria related to safety systems and pressure boundaries. The safety system is a general term for equipment that, when an abnormal state such as an overpower state or an abnormal increase in power of the reactor P1 is detected during reactor operation, activates the reactor emergency shutdown system to automatically (or manually) shut down the reactor P1, and activates engineered safety facilities to protect the core P13 and the containment vessel boundary in the event of an accident. The pressure boundary refers to the primary reactor system, that is, the pressure vessel P12 through which high-temperature and high-pressure coolant circulates, and the main cooling pipe system.

[0045] Selection in the selection step S1 (risk judgment) may use various risk analysis tools (e.g., stress corrosion cracking crack growth rate analysis tools) or judgment by experts (expert meeting). This allows high-risk equipment or parts to be prioritized in terms of the importance of SCC protection and targeted for the ER process. For example, in a BWR, equipment and structures that are difficult to access for repair or replacement and that form a pressure boundary include the control rod drive mechanism housing at the bottom of the reactor, the neutron instrumentation tube housing, and the measurement nozzle. Considering the materials used, stress, and water quality in these parts, they are high-risk parts. Similar parts that do not form a pressure boundary but are important in terms of flooding of the core P13 and retention of fuel in the event of an accident include the weld lines of the jet pump P21 diffuser, the jet pump P21 riser, the shroud support P41, the neutron instrumentation tube guide tube, and the shroud P15. Items that can be repaired or replaced include the recirculation system piping P30, the inlet and outlet nozzle safe ends of the recirculation system piping P30, the measurement nozzle safe ends, the core spray piping, the dryer, the separator, and other equipment, the inlet mixer of the jet pump P21, the jet pump beam of the jet pump P21, and the shroud head bolts of the shroud P15. SSCs with even lower risks include areas that are relatively easy to access and work on, such as the top of the reactor. In the selection step S1, the risks of such SSCs are classified, priorities are set for improving reliability, and the areas to be monitored are selected. In this improvement method, the areas to be monitored in the monitoring step S2 are determined based on the results selected in the selection step S1. In the monitoring step S2, which receives the selection results, the process described below is carried out.

[0046] In this improvement method, after selection step S1, monitoring step S2 is performed. In monitoring step S2, the ECP of the high-risk equipment or parts selected in selection step S1 is monitored (measured) as a constant monitoring index, and the crack growth rate is estimated from the results (performance monitoring, crack evaluation). Here, the constantly monitored ECP is evaluated for the SCC crack growth rate using plant data from the design and construction of materials, stress, etc., as well as water quality parameters other than ECP. ECP makes it possible to see the corrosive environment of the target part during operation, making it possible to evaluate and estimate the crack growth rate of the monitored part.

[0047] In this improvement method, the state in which SCC is managed during the operation of the BWR is considered to be the "ideal performance" of the BWR with respect to SCC, and the ECP is constantly monitoring the occurrence and growth of SCC as a suitable monitoring index, and the crack growth rate evaluated by the ECP is defined as an essential condition for improving the ER. The corrosion resistance and stress of the material also affect SCC, but these cannot be constantly monitored as performance during reactor operation. Since the corrosion resistance and stress state of the material must be measured during reactor shutdown, they are not directly used as monitoring indexes in this embodiment. However, the measured values ​​of the corrosion resistance and stress of these materials are stored in a database for evaluating the crack growth rate, and can be used to evaluate the crack growth rate together with the actual measured values ​​of the ECP. In addition, the reactor water electrical conductivity due to impurity ions also affects the crack growth rate, so the measured values ​​can be used. The reactor water electrical conductivity is already constantly monitored by the main equipment of the reactor water. In addition, the inspection results regarding the cracks are also stored in a database and used to calibrate the evaluated crack growth rate.

[0048] In this improvement method, after the monitoring step S2, a preventive maintenance execution step S3 is performed. In the preventive maintenance execution step S3, the presence or absence of cracks in the target area is confirmed and the crack length is measured during a regular inspection according to an inspection plan. In this improvement method, based on the inspection results and the crack growth rate evaluated in the monitoring step S2, it is determined whether or not operation can be continued with the crack remaining and when the next inspection should be scheduled.

[0049] In determining the timing of the next inspection, this preventive maintenance execution step S3 uses the ratio of the continuously monitored crack growth rate to the crack growth rate estimated for equipment or parts with a high SCC risk when continuous monitoring is not performed, as described above. Here, if no cracks are found in equipment or parts with high maintenance importance, the preventive maintenance execution step S3 can determine the next inspection time to be extended by a factor of the above-mentioned ratio. In addition, if a crack is found in equipment or an area of ​​high maintenance importance, and the length of the found crack is greater than the estimated crack growth rate that is constantly monitored, a decision can be made in the preventive maintenance execution step S3 to conduct an inspection every time. Furthermore, if a crack is found in equipment or a part that is highly important for maintenance, and the length of the found crack is equal to or less than the estimated crack growth rate that is constantly monitored, the preventive maintenance execution step S3 can decide to conduct inspections every other minute. When determining the next inspection time, if the decision is made based on these case distinctions in the preventive maintenance execution step S3, the inspection frequency can be reduced while improving the ER of the plant.

[0050] In this improvement method, it is preferable to carry out an improvement step S4 after the preventive maintenance execution step S3. In the improvement step S4, repairs or replacements are carried out for equipment or parts that are determined in the preventive maintenance execution step S3 to have a crack length that will soon exceed the allowable value. In addition, in the improvement step S4, preventive maintenance such as stress improvement is carried out at the implemented location (by applying preventive maintenance technology) to prevent the occurrence and growth of similar cracks. As a result, improvements are made in a planned manner for the found cracks, so the reliability of the equipment or parts selected as having high maintenance importance in the selection step S1 is continuously improved. As a result, the SCC risk of the reactor P1 can be reduced as a whole and the ER can be improved. Now, in this improvement method, a case where the corrosive environment mitigation technology is not applied is considered, so it is preferable to carry out material replacement or stress improvement in the improvement step S4 for the equipment or parts selected in the selection step S1 as having high maintenance importance. For example, the material is replaced with a material with higher corrosion resistance. For example, material replacement involves changing from 182 alloy to improved nickel-based alloy weld metal 182 or 82 alloy, or from 304 stainless steel to low carbon 316 stainless steel, and piping and equipment replacement and weld overlay are also applied. Stress improvement involves applying compressive stress to the surface by peening such as water jet peening and shot peening, surface polishing, or mechanical stress improvement (MSIP). By applying strong compressive stress to the surface of the material by peening, it is possible to prevent the occurrence of SCC for a long period of time, such as several decades. After peening, the initial inspection period is equivalent to that of a new plant.

[0051] In this improvement method, after the improvement step S4, a life cycle management step S5 is performed. In the life cycle management step S5, a maintenance plan is formulated for repeatedly performing the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, and the improvement step S4, and the plan is executed. In addition, in the life cycle management step S5, the ECP of the equipment or part selected as high risk in the selection step S1 is monitored to manage the crack growth rate. In the life cycle management step S5, a plan is formulated and executed for repeatedly performing this series of procedures in the medium to long term. That is, in the life cycle management step S5, each step in the above-mentioned ER process is systematically and repeatedly performed, and the ER of the plant is continuously improved in the medium to long term. This makes it possible to reduce the overall SSC risk of the nuclear plant P100 to ALARP. In addition, in the life cycle management step S5, based on the SCC database created based on the ECP data, inspection data, and SCC occurrence and progression data from other plants accumulated through the long-term implementation of the ER process consisting of the above steps, the time interval for checking the SCC state by SCC inspections performed for preventive maintenance can be extended, and the SCC crack growth rate can be managed by replacing the period with ECP monitoring. Furthermore, in the life cycle management step S5, for equipment or parts that have been constantly monitored, including those that have had cracks progress as evaluated by inspection or where no cracks were found, the inspection interval can be extended on the premise of constant ECP monitoring. This improvement method allows for the improvement and rationalization of ER to proceed at the same time. This also makes it possible to review regular inspection items. This is a feature brought about by the ER process in this improvement method.

[0052] <<Specific example of ER process in this improvement method>> Next, a specific example of the ER process in this improvement method will be described with reference to Figures 1 and 4. Figure 4 is an explanatory diagram showing an example of the effect obtained by this improvement method (a graph of ER improvement by constant monitoring of crack growth rate) over time. The horizontal axis in the figure indicates the operation cycle (time), the vertical axis on the left side indicates the crack growth rate (mm / s), and the vertical axis on the right side indicates ECP (V vs. SHE) and electrical conductivity (μS / m). The crack growth rate shown by the solid line in the figure is read on the left vertical axis, and the ECP shown by the short dashed line and the electrical conductivity shown by the long dashed line are read on the right vertical axis.

[0053] As shown in Fig. 1, first, in the selection step S1, the equipment or parts are ranked in terms of the importance of SSC maintenance (classification of equipment or parts). In the selection step S1, the risk ranking regarding SCC of the equipment / structure or its parts (hereinafter, sometimes simply referred to as "parts") is determined. In the selection step S1, the parts with the highest risk of SCC are selected as the ECP constant monitoring target (i.e., the part to be constantly monitored for crack growth rate) in the monitoring step S2. The part to be constantly monitored for crack growth rate is a place with a high SCC risk, that is, a place with a high crack growth rate and a short time (allowable time) for SCC cracks to penetrate or maintain the bond cross section required for strength, and therefore has at least one characteristic of a place where the corrosion resistance of the material is reduced or the stress conditions are severe, or a place with a significantly high ECP. When the part with a high crack growth rate is constantly monitored, it can be considered that the crack growth rate is lower than the crack growth rate constantly monitored for other parts with a low crack growth rate and a relatively long allowable time without directly monitoring the ECP and constantly evaluating the crack growth rate. Therefore, the number of ECP sensors to be installed for constant monitoring can be reduced. In addition, the installation of ECP sensors in locations where it is difficult to install them is not required, which simplifies the process. The determination of whether a location has a high or low crack growth rate can be determined from the viewpoint of materials and stress, taking into account the materials used, processing and manufacturing methods, etc., from data at the time of design and construction. In particular, it may be determined based on the results of evaluations using various crack growth evaluation formulas and analytical models. In addition, any ECP sensor that functions when immersed in the reactor water conditions of a BWR can be used as the ECP sensor required at this time.

[0054] Then, in the monitoring step S2, the ECP of the high-risk parts selected in the selection step S1 is constantly monitored as a monitoring index. In addition, in the monitoring step S2, the crack growth rate is estimated using the monitoring results, water quality parameters other than the ECP, and a plant database of materials, stresses, etc. Figure 4 specifically shows the state of the performance index in the monitoring step S2 over time, and clearly shows the output of the ECP sensor during a given operating cycle.

[0055] As mentioned above, the ECP at normal water quality is 0 mV vs. SHE or more, and in many cases 100 mV vs. SHE or more. The ECP changes when the core flow rate or the control rod pattern change due to the management of the core P13. The cause of the change is that the radiation field of the core P13 or the amount of water present in the core P13 changes due to the operation management of the core P13, and as a result, the concentration of oxygen and hydrogen peroxide generated by radiolysis of the reactor water changes. In addition, the electrical conductivity of the reactor water is continuously monitored in the sampling system of the BWR. In the pure water management of the BWR, the value is usually 10 μS / m or less in many plants, but the electrical conductivity may temporarily increase due to the introduction of impurities from the feedwater. The electrical conductivity may also increase slowly due to the deterioration of the resin in the purification system. In such cases, appropriate measures are taken promptly in accordance with the water quality management standards, and the water quality is restored. If the cause of such an increase in electrical conductivity is chloride ions or sulfate ions that have a negative effect on SCC, it will cause an increase in the crack propagation rate, as shown in Figure 4. In this way, by constantly monitoring the ECP and electrical conductivity, when a crack occurs during a regular inspection, it is possible to know for the first time what kind of environment the crack has been exposed to, and to compare it with the measured value when evaluating the crack length. The crack growth rate during the operating cycle in the figure is evaluated using the constantly monitored ECP and electrical conductivity, and the crack length is given as the integral value of the crack growth rate at every moment. Then, it becomes possible to compare it with the crack length actually found by inspection during a regular inspection. By operating while constantly monitoring that the crack growth rate evaluated from the ECP value is within the range of the control target, it is expected that unexpected crack growth will not occur. In this way, by constantly monitoring the ECP, calculating the crack growth rate by combining it with other plant information, and constantly monitoring the crack growth rate, SCC cracks in BWRs can be controlled, that is, it is possible to improve the ER by keeping the performance index in a controlled state.

[0056] When an inspection is performed in the preventive maintenance execution step S3 in Figures 1 and 2, the presence or absence of cracks is confirmed. If a crack is found, the length of the crack is measured, and the crack growth rate is calculated to evaluate the impact on operation. The setting of the next inspection time at this time will be explained using Figure 5.

[0057] FIG. 5 is a diagram showing an example of the values ​​of the monitoring indexes over time in a nuclear power plant P100 to which this improvement method is applied. The horizontal axis of the figure shows the time (operation cycle) (times) of each regular inspection. The vertical axis shows the crack propagation amount (mm). As shown in the graph on the left side of FIG. 5, when no cracks are found in the target part during the regular inspection, the crack propagation amount during that regular inspection is 0 mm. Therefore, when no cracks are found in the target part during the regular inspection, the initial value is set to 0 mm (or the initial crack length required for evaluation), and the crack propagation rate is calculated using an evaluation formula of an academic society or an appropriate evaluation model to evaluate the future crack length. At this time, ECP is not generally measured in a BWR with normal water quality to which no corrosive environment mitigation technology is applied. Therefore, the calculated (estimated) crack propagation rate is calculated using the value of ECP at normal water quality. For example, in the Japan Society of Mechanical Engineers' Standard for Maintenance of Nuclear Power Plant Equipment for Power Generation (hereinafter referred to as the Maintenance Standard), the ECP for normal water quality is set to 150 mV vs. SHE, and an evaluation formula for the crack propagation rate at that time is given. Therefore, by assuming a constant ECP during one operating cycle, the crack growth rate is determined by the stress, material and initial crack length at the site.

[0058] In addition, in the maintenance standards, the crack growth rates for SUS316L for piping and reactor internals in a BWR at normal water quality and when hydrogen is injected are given by the following equations.

[0059] Relationship between crack growth rate da / dt (m / s) and stress intensity factor K (MPa√m) in normal water quality (ECP≧150mV vs.SHE and electrical conductivity<20μS / m): da / dt=0 (K≦0) da / dt=3.33×10-14 K 2.161 (0 <K≦57.9) da / dt=2.1×10 -10 (K>57.9)

[0060] Relationship between crack growth rate da / dt (m / s) and stress intensity factor K (MPa√m) during hydrogen injection (ECP≦-100mV vs. SHE and electrical conductivity <20μS / m): da / dt=0 (K≦0) da / dt=3.33×10 -15 K 2.161 (0 <K≦57.9) da / dt=2.1×10 -11 (K>57.9)

[0061] The following is also given for sensitized SUS304 for piping and reactor internals:

[0062] Relationship between crack growth rate da / dt (m / s) and stress intensity factor K (MPa√m) in normal water quality (ECP≧150mV vs.SHE and electrical conductivity<20μS / m): da / dt=0 (K≦0) da / dt=1.43×10 -13 K 2.161 (0 <K≦57.9) da / dt=9.2×10 -10 (K>57.9)

[0063] Relationship between crack growth rate da / dt (m / s) and stress intensity factor K (MPa√m) during hydrogen injection (ECP≦-100mV vs. SHE and electrical conductivity <20μS / m): da / dt=0 (K≦0) da / dt=7.94×10 -15 K 2.161 (0 <K≦57.9) da / dt=5.1×10 -11 (K>57.9)

[0064] On the other hand, when ECP is constantly monitored even under normal water quality as in this improvement method, the conservatism during evaluation can be rationally reduced by using the actual measured value from constant monitoring as the ECP used to evaluate the crack growth rate. For example, the maintenance standard also gives the crack growth rate when hydrogen is injected, and this can be applied when the ECP is -100 mV vs. SHE or lower. The crack growth rate for ECP between the ECP under normal water quality and the ECP under hydrogen injection is not given for stainless steel. The crack growth rate between them is calculated by apportioning it with the actual measured value of ECP, so the crack growth rate C can be evaluated from the ECP obtained by constant monitoring under normal water quality. Therefore, when the ECP value obtained by constant monitoring is between -100 and 150 mV vs. SHE, in 304 stainless steel and 316L stainless steel, if the ECP drops by about 150 mV, the crack growth rate C will be lower than the crack growth rate C under normal water quality. N Therefore, the formula for apportionment is given by the following Equation 1.

[0065]

number

[0066] As a result, for example, if +100mV vs. SHE is the actual measured value, the evaluation value for the crack growth rate (A in Fig. 5) will be about half the crack growth rate at normal water quality (150mV vs. SHE) (B in Fig. 5) specified by the maintenance standard. Note that the straight line in Fig. 5A shows the crack growth rate evaluated from monitoring the crack growth rate. The dashed line in Fig. 5B shows the crack growth rate evaluated without monitoring the crack growth rate.

[0067] Therefore, the crack growth rate under normal water quality evaluated without measuring ECP was set as C NE The crack growth rate evaluated by this improvement method is C E Then, the conservativeness R that can be reasonably reduced is M teeth, R M =C NE / C E …(Formula 2) Therefore, as mentioned above, if the measured ECP is 100mV vs. SHE as a cycle average, and it is 50mV lower than the ECP of normal water quality given by the maintenance standard of 150mV vs. SHE, then R M is approximately 2, so the next inspection time can be extended by twice that ratio (in this case, 2 times) (C in Figure 5). However, in the next and subsequent operating cycles, it is necessary to keep the crack growth rate under control by constant monitoring of the ECP, and to operate while confirming that the crack growth rate evaluated from the ECP is at the same level. In the case of nickel alloy weld metal 182, crack growth rates are given for ECP ≧ -100mV vs.SHE or more, -200mV vs.SHE ≦ ECP ≦ -100mV vs.SHE, and ECP ≦ -200mV vs.SHE.

[0068] The graph on the right side of Figure 5 shows the setting for the next inspection when cracks are found. Unlike the graph on the left side of Figure 5 (when no cracks are found), the starting point for the evaluation of crack length and crack growth rate is the length of the crack discovered by inspection. In this case, materials with low corrosion resistance remain in old plants, and if cracks are found in such areas, inspection will be required at each regular inspection (multiple "●" in the graph on the right side of Figure 5). However, in the case of this improvement method, which constantly monitors ECP to evaluate crack growth rate, as described above, if the cycle average ECP is, for example, 100mV vs.SHE, which is lower than the ECP of 150mV vs.SHE for normal water quality as specified in the maintenance standard, then the R given by Equation 2 will be M is approximately 2, and inspections can be performed at alternate times. In other words, for example, inspections can be exempted on the occasions where the "●" and the "dashed circle" overlap in the graph on the right side of Figure 5 (D in Figure 5). Even in this case, the crack growth rate is constantly monitored by the ECP using this improvement method, and it is ensured that no particular increase in the crack growth rate is observed on the cycle average, improving ER and allowing for a reasonable maintainability R without compromising safety. M It is possible to reduce

[0069] In this way, the RM If the crack growth rate itself is high and it is expected to reach the allowable time even if ER is reduced, the improvement step S4 in Fig. 1 is performed. In the improvement step S4, at least one of material replacement and repair is performed, and further, stress improvement such as peening is performed as a preventive maintenance measure to reduce the risk of similar cracks occurring again, and ER is continuously improved.

[0070] Then, a mid- to long-term maintenance plan is formulated and executed by the life cycle management step S5 shown in FIG. 1. That is, the life cycle management step S5 repeats the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, and the improvement step S4 during the plant operation period. The life cycle management step S5 monitors the ECP of the part selected in the selection step S1 as being at high risk in terms of the importance of SCC maintenance, and continues to manage the crack growth rate. That is, in this improvement method, by continuously implementing these steps as an ER process, the risk reduction of the SCC of the entire plant is ALARP. The part with the highest risk in the early stage of applying this improvement method will have its risk reduced by implementing the improvement step S4, and the high-risk part will transition to a different part as the operation years progress. Therefore, by continuing to clearly indicate the part with a high risk in a certain operation cycle by constant monitoring of the ECP, it can be shown that the risk of other parts is also maintained low. In this way, in this improvement method, the ER of the BWR plant is continuously improved in the mid- to long-term by implementing the ER process described above. Since the crack growth rate at the highest risk part in the relevant operating cycle is monitored, this improvement method also determines the next inspection time for other low-risk parts based on the R evaluated at the highest risk part. M This improvement method reduces the frequency of inspections by constantly monitoring ECP, especially in situations where the above-mentioned corrosive environment mitigation techniques are not being implemented.

[0071] <<Variations>> In the monitoring step S2, at least one water quality parameter selected from the oxygen concentration of reactor water, hydrogen peroxide concentration of reactor water, and electrical conductivity of reactor water, the correlation of which with the ECP monitored at all times is periodically obtained in advance, can be used as an auxiliary constant monitoring index. This makes it possible to guarantee the ER in places where a corrosion potential sensor for monitoring the ECP cannot be installed, or during periods when the corrosion potential sensor cannot be used temporarily due to a malfunction or the like. Note that the relationship between the ECP and the above auxiliary constant monitoring index is affected by the state of the power control of the nuclear fuel installed in the core P13 of the reactor P1, and changes depending on the burnup of the nuclear fuel, the position of the control rod, the core flow rate, etc., so it is difficult to continue using the correlation for a long period of time. Therefore, it is preferable to periodically re-acquire these relationships at each fuel exchange, or at most every five years when all the fuel is replaced.

[0072] Furthermore, as described above, the present improvement method can be applied not only to BWRs but also to ABWRs. As described above, Fig. 3 shows an example of the case where the present improvement method is applied to a nuclear power plant P100 using a BWR. In the nuclear power plant P100 using a BWR, reactor water from the lower part of the downcomer P17 flows through the recirculation system piping P30 and the purification system piping P20. As described with reference to Fig. 3, in the case of the nuclear power plant P100 using a BWR, the ECP is measured by a corrosion potential sensor P35a mounted on a flange P36 provided on a branch line P34a provided from a bottom drain line P34.

[0073] In this regard, in the case of a nuclear power plant P100 using an ABWR, the ECP can be measured as shown in Fig. 6. Fig. 6 is an overall system configuration diagram showing another example of a nuclear power plant to which this improvement method is applied (a nuclear power plant P100 using an ABWR). The nuclear power plant P100 shown in FIG. 6 differs from the nuclear power plant P100 shown in FIG. 3 in the following ways. The nuclear power plant P100 shown in FIG. 6 does not include a recirculation system piping P30 and a jet pump P21. The nuclear power plant P100 shown in FIG. 6 includes an internal pump P40 at the bottom of the pressure vessel P12. In the nuclear power plant P100 shown in FIG. 6, an extraction piping P14 is connected between the turbine P3 and the low-pressure feedwater heater P8 and the high-pressure feedwater heater P9. Extraction steam extracted from the main steam piping P2 and the turbine P3 by the extraction piping P14 is supplied to the low-pressure feedwater heater P8 and the high-pressure feedwater heater P9 by the extraction piping P14, respectively, and serves as a heating source for the feedwater. The drain of the extraction steam from the low-pressure feedwater heater P8 is returned to the condensate pump P5. The drain of the extraction steam from the high-pressure feedwater heater P9 is returned to the feedwater piping P10.

[0074] As shown in Fig. 6, in the case of a nuclear power plant P100 using an ABWR, the ECP is measured in the reactor water above the downcomer P17. To realize this, in the case of the nuclear power plant P100 using an ABWR, a pipe P20a connected to a purification system pipe P20 branched off from a residual heat removal system (not shown) connected above the downcomer P17 can be used. Specifically, as shown in Fig. 6, a flange P36 is installed on the pipe P20a, and an ECP sensor P35e is loaded thereon to measure the ECP. Although not shown, in the nuclear power plant P100 using an ABWR, a branch line is provided on the pipe P20a, a flange is installed on the branch line, and an ECP sensor is loaded thereon to measure the ECP. [Explanation of symbols]

[0075] S1 Selection Steps S2 Monitoring Steps S3 Preventive maintenance execution steps S4 Improvement steps S5 Life Cycle Management Steps S6 Corrective steps P100 Nuclear Plant P1 reactor P2 Main steam pipe P3 Turbine P4 Condenser P5 Condensate Pump P6 Condensate purification equipment P7 Water Pump P8 Low pressure feed water heater P9 High pressure feed water heater P10 Water supply piping P11 Nuclear containment vessel P12 Pressure vessel P13 Core P14 Extraction piping P15 Shroud P17 Down Cam P20 Purification system piping P20a piping P21 Jet pump P23 Purification system isolation valve P24 Purification system pump P25 Regenerative heat exchanger P26 Non-regenerative heat exchanger P27 Reactor water purification equipment P30 Recirculation system piping P34 Bottom drain line P34a Branch Line P35a~P35e Corrosion Potential Sensor P36 flange P38 Neutron Instrumentation Tube P40 Internal Pump P41 Shroud Support

Claims

1. A selection step of selecting equipment or parts with a high risk of stress corrosion cracking as equipment or parts with high conservation importance; a monitoring step of monitoring the corrosion potential of the selected equipment or portion having high conservation importance as a monitoring index, and constantly monitoring the crack growth rate at the monitored portion during an operation cycle using the corrosion potential based on water quality parameters and plant data; a preventive maintenance execution step of determining the next inspection timing for preventive maintenance using a ratio of the constantly monitored crack growth rate and a crack growth rate estimated in the equipment or part having a high risk of stress corrosion cracking when the constant monitoring is not performed; 1. A method for improving reliability of a nuclear power plant, comprising:

2. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the selection of the equipment or portion with high importance for maintenance is made based on at least one of a criterion related to a safety system for at least one of a structure, a system, and an equipment of the nuclear power plant and a criterion related to a pressure boundary, by at least one of a crack growth rate analysis tool for stress corrosion cracking and a meeting of experts.

3. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the monitoring of the crack growth rate comprises estimating the crack growth rate using plant data of the monitored corrosion potential, electrical conductivity, impurity ion concentration, and material and stress.

4. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the continuous monitoring of the crack growth rate is performed in combination with at least one of a reactor water oxygen concentration, a reactor water hydrogen peroxide concentration, and a reactor water electrical conductivity, the correlation of which with the continuously monitored corrosion potential being periodically determined in advance.

5. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the estimated crack growth rate is calculated using a corrosion potential value in normal water quality.

6. After the preventive maintenance execution step, 2. The method for improving reliability of a nuclear power plant according to claim 1, further comprising an improvement step of improving equipment reliability by carrying out replacement, repair, and preventive maintenance of equipment or parts determined to have a crack length that will soon exceed an allowable value.

7. After the improvement step, 7. The method for improving reliability of a nuclear plant according to claim 6, further comprising a life cycle management step of formulating a maintenance plan for repeatedly performing said selecting step, said monitoring step, said preventive maintenance execution step, and said improvement step over a medium to long term, and executing said maintenance plan.

8. When no cracks are found in the equipment or part with high maintenance importance, the preventive maintenance execution step can determine the next inspection time to be extended by a factor of the ratio; When a crack is confirmed in the equipment or part with high maintenance importance, if the length of the confirmed crack is greater than the evaluation of the crack growth rate monitored at all times, the preventive maintenance execution step decides to perform an inspection every time, and if the length of the confirmed crack is equal to or less than the evaluation of the crack growth rate monitored at all times, the preventive maintenance execution step decides to perform an inspection every other time.

2. The method for improving reliability of a nuclear power plant according to claim 1.

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