Methods for improving nuclear plant reliability.

Abstract

To provide a method for improving the reliability of a nuclear power plant, in which corrosion potential can be used as a constant monitoring index as a performance index in a process of reliability-oriented maintenance, and thereby the reliability of the nuclear power plant can be improved.SOLUTION: A method for improving the reliability of a nuclear power plant according to the present invention includes: a selection step S1 of selecting a portion or the like with a high risk of stress corrosion cracking; a monitoring step S2 of monitoring corrosion potential of the selected portion and estimating a crack growth rate; a preventive maintenance execution step S3 of reducing the corrosion potential at a portion or the like where the estimated crack growth rate is high; a correction step S4 of correcting water quality parameters so that the corrosion potential falls within a target range; an improvement step S5 of improving the reliability by replacing materials or the like; and a life cycle management step S6 of monitoring the corrosion potential and managing the crack growth rate while repeatedly performing the-above steps.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 referred to as nuclear plants) are made of structural materials such as stainless steel and nickel-based alloys. These structural materials are susceptible to stress corrosion cracking (hereafter referred to as SCC) under certain conditions. Therefore, SCC prevention measures are applied to maintain the integrity of nuclear reactors. In recent years, SCC prevention measures have also been applied from the perspective of improving economic efficiency, such as increasing the capacity factor of nuclear reactors, and from the perspective of responding to the aging of nuclear plants.

[0003] Measures to prevent SCC are being applied to improve the corrosion resistance of materials, improve stress, or mitigate the corrosive environment. In boiling water reactors (BWRs), hydrogen injection is widely used both domestically and internationally as one of the SCC countermeasures based on improving the corrosive environment of the cooling water (reactor water) in the pressure vessel to which structural components are exposed. Reactor water contains oxygen and hydrogen peroxide, which are generated by radiolysis of the cooling water in the pressure vessel and cause corrosion. Reactor water containing oxygen and hydrogen peroxide forms a corrosive environment.

[0004] The corrosion potential (ECP) of structural materials is used as an indicator of the corrosive environment. It is known that SCC is suppressed when the ECP is lower than -300 to -200 mV vs. SHE.

[0005] Hydrogen injection is a technology that adds hydrogen to reactor water by supplying hydrogen-injected feedwater into the pressure vessel, and then reacts this hydrogen with the oxygen and hydrogen peroxide contained in the reactor water to return it to water, thereby reducing ECP and suppressing the occurrence of SCC.

[0006] Another known technology for promoting the reduction of ECP during hydrogen injection is the injection of platinum group metal elements into the reactor water. This technology utilizes the catalytic action of the injected platinum group metal elements on the electrochemical reaction of hydrogen to further reduce ECP during hydrogen injection.

[0007] In these conventional techniques, it is necessary to accurately know the ECP of the structural material in order to evaluate the SCC suppression effect. Therefore, corrosion potential sensors are installed inside the pressure vessel or on the piping connected to the pressure vessel to measure the ECP of the structural material.

[0008] While ECP is the primary factor influencing SCC, it is also strongly affected by impurity ions, particularly chloride and sulfate ions, present in the reactor water. In other words, even for a given ECP, the higher the impurity concentration in the reactor water, the more likely SCC will occur. Therefore, SCC susceptibility (including information on the initiation and propagation rate of cracks) can be said to depend on ECP and electrical conductivity. Electrical conductivity is determined by the impurity concentration in the reactor water. For this reason, BWR nuclear power plants have been operating to maintain high reactor water purity by reducing impurity concentrations even before lowering ECP. On the other hand, if ECP is sufficiently low, for example, at around -400 mV vs. SHE, the impact of increased electrical conductivity on SCC remains within a certain margin. It has also been shown that even if the electrical conductivity due to impurity ions increases from the usual 0.1 μS / cm to around 0.2 μS / cm, SCC susceptibility is small. Therefore, noble metal injection, which can maintain a significantly lower ECP, is an advantageous technology for SCC management in nuclear power plants.

[0009] Under these circumstances, a technique for monitoring SCC using ECP and evaluating crack length taking into account the time to SCC onset is known (see, for example, Patent Document 1). Patent Document 1 describes a method for measuring the ECP of a structural member that comes into contact with cooling water containing impurity ions and the concentration of impurity ions contained in the cooling water, determining the time to SCC onset in the structural member based on this information, and reducing at least one of the ECP and the concentration of impurity ions if this onset time is shorter than a set time.

[0010] Furthermore, a life prediction method for a primary reactor structure that predicts crack propagation due to SCC and crack propagation due to corrosion fatigue is known (see, for example, Patent Document 2). The life prediction described in Patent Document 2 is performed using measured ECP and crack propagation characteristic data of structural materials.

[0011] Other methods for estimating the remaining life of structural components in 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 within a nuclear reactor for which the remaining life is to be estimated. This method then uses these values ​​to determine the growth rate of a crack due to stress corrosion for a crack assumed to exist in the portion for which the remaining life is to be estimated, and calculates the time it takes for the crack to reach a predetermined limit value from the crack growth rate, which is then used to determine the remaining life. [Prior art documents] [Patent documents]

[0012] [Patent Document 1] Patent No. 5483385 [Patent Document 2] Japanese Patent Application Laid-Open No. 2006-10428 [Patent Document 3] Japanese Patent Application Publication No. 6-34786 Summary of the Invention [Problem to be solved by the invention]

[0013] After evaluation, prediction, estimation, etc. are performed as described in Patent Documents 1 to 3, various countermeasures are applied as necessary. For example, countermeasure technologies including stress improvement, including corrosive environment mitigation technology for SCC, material replacement, and SCC crack inspection technology are applied as measures to prevent the occurrence and progression of SCC, repair discovered SCC cracks, and post-repair measures. Conventionally, these have mainly been applied individually. Furthermore, when operating while allowing cracks, it has been necessary to continuously check the state of the cracks through inspection.

[0014] Meanwhile, the introduction of reliability-centered maintenance (RCM) is being promoted worldwide in nuclear power plants. A representative example of this is AP913 (Equipment Reliability Process) from the American Nuclear Power Organization (INPO). RCM involves formulating an optimal maintenance program to improve equipment reliability by selecting critical structures, systems, and components (hereafter, collectively referred to as "SSCs"), optimal maintenance methods, and implementation timing. The introduction of RCM should aim to improve the reliability of the entire nuclear plant. However, currently, there is no RCM process for managing SCC, which has a significant impact on the reliability of nuclear plants. Therefore, as mentioned above, SCC prevention technologies are limited to being applied individually.

[0015] By linking SCC management with RCM, it is possible to link individual countermeasures and SCC management procedures, and to establish a process for long-term and continuous reliability improvement throughout the entire lifecycle of a nuclear power plant. This is expected to further improve the reliability of nuclear power plants. In the RCM process, specific performance indicators must be constantly monitored and clearly indicated to indicate the state of equipment reliability, but since there was no RCM process for SCC, constant monitoring indicators had not been established.

[0016] 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 for a reliability-based maintenance (RCM) process and can continuously monitor the ECP, thereby improving the reliability of the nuclear power plant. [Means for solving the problem]

[0017] In order to solve the above problems, the present invention embodies an equipment reliability process for SCC, in which ECP is used as a performance index for constantly monitoring and indicating equipment reliability against SCC, and the SCC crack growth rate (amount of crack growth) is evaluated over the plant lifecycle using the monitored ECP values ​​as input along with the stress state of the materials used in each SSC and the welds.

[0018] Specifically, the present invention includes a selection step of selecting equipment or parts at high risk of stress corrosion cracking, a monitoring step of constantly monitoring the corrosion potential of the selected high-risk equipment or parts as a monitoring index and estimating a crack growth rate from the result, a preventive maintenance execution step of performing preventive maintenance to reduce the corrosion potential of the equipment or parts at high risk, a correction step of correcting water quality parameters so that the corrosion potential monitored in the monitoring step falls within a target range, and an improvement step of, if the corrosion potential does not fall within the target range even after the correction step, performing at least one of material replacement and stress improvement to continuously improve the reliability of the equipment or parts selected as high risk in the selection step, and a life cycle management step of repeatedly performing the selection step, the monitoring step, the preventive maintenance execution step, the correction step, and the improvement step, while monitoring the corrosion potential of equipment or parts newly selected as high risk in the repeated selection step, and managing the crack growth rate. The life cycle management step extends the time interval for checking the state of stress corrosion cracking by inspection based on a stress corrosion crack database constructed based on at least one of corrosion potential data, inspection data, and stress corrosion cracking occurrence and progression data in other plants, which are accumulated in the process of repeatedly performing the selection step, the monitoring step, the preventive maintenance execution step, the correction step, and the improvement step, and replaces the time interval with monitoring the corrosion potential to perform stress corrosion cracking growth rate management. The present invention relates to a method for improving the reliability of a nuclear power plant. [Effects of the Invention]

[0019] According to the present invention, a method for improving the reliability of a nuclear power plant can be provided, in which the corrosion potential (ECP) can be used as a performance index for a reliability-based maintenance (RCM) process and can be continuously monitored, 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 explanation of the drawings]

[0020] [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. [Figure 2] 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 3] 1 is an explanatory diagram showing an example of an effect (a graph of risk reduction) obtained by a method for improving reliability of a nuclear power plant according to an embodiment of the present invention over time. [Figure 4] 1 is an explanatory diagram showing an example of an effect (a graph of risk reduction) obtained by applying a method for improving reliability of a nuclear power plant according to an embodiment of the present invention to a boiling water reactor over time. [Figure 5] 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 INVENTION

[0021] Hereinafter, a method for improving reliability of a nuclear power plant according to one embodiment of the present invention (hereinafter, sometimes 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 will be denoted by the same reference numerals, and duplicated descriptions may be omitted.

[0022] ≪This improvement method≫ FIG. 1 is a block diagram showing the functional flow of the present improvement method. As shown in FIG. 1, this improvement method includes a selection step S1, a monitoring step S2, a preventive maintenance execution step S3, a correction step S4, an improvement step S5, and a life cycle management step S6.

[0023] In the selection step S1, equipment or parts with a high risk of stress corrosion cracking (SCC) are selected. In the monitoring step S2, the corrosion potential (ECP) of the selected high-risk equipment or parts is constantly monitored as a monitoring index, and the crack growth rate is estimated from the result. In the preventive maintenance execution step S3, preventive maintenance is performed to reduce the ECP of the equipment or parts having a high estimated crack growth rate. In the correction step S4, the water quality parameter is corrected so that the ECP monitored in the monitoring step S2 falls within the target range (that is, the water quality is controlled so that the water quality parameter is corrected). In the improvement step S5, if the ECP does not fall within the target range even after the correction step S4 is performed, at least one of material replacement and stress improvement is performed to continuously improve the reliability of the equipment or parts selected as high risk in the selection step S1. In the life cycle management step S6, the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, the corrective step S4, and the improvement step S5 are repeatedly performed. Also, in the life cycle management step S6, the ECP of equipment or parts newly selected as high risk in the repeated selection step S1 is monitored to manage the crack growth rate. The monitoring step S2 can be carried out as appropriate after the preventive maintenance execution step S3, the corrective step S4, and the improvement step S5 have been carried out.

[0024] In this improvement method, countermeasures are implemented in the preventive maintenance execution step S3, the corrective action step S4, and the improvement step S5. Therefore, by repeatedly performing this equipment reliability process for the nuclear power plant, the risk of the high-priority (high-risk) equipment or components selected in the selection step S1 decreases over time, and the risk ranking of each SSC changes. In other words, by repeating the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, the corrective action step S4, and the improvement step S5, high-risk equipment or components are constantly monitored and countermeasures are continuously implemented. Therefore, this improvement method reduces the overall risk of the nuclear power plant's SSCs to a reasonably feasible level (ALARP). In this improvement method, continuous monitoring of the ECP consistently demonstrates that the ECP is controlled to fall within the target range, thereby reducing the SCC risk to ALARP. This demonstrates that each SSC in configuration management is performing as designed and performing as expected, demonstrating good equipment reliability.

[0025] Examples of nuclear reactors to which this 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-cooled reactors, heavy water-moderated reactors, molten salt reactors, and fast breeder reactors. An example of a configuration in which this improvement method is applied to a nuclear power plant (nuclear plant) using a BWR will be described below. Specific details of this improvement method will then be described based on this example configuration. Figure 2 is an overall system configuration diagram showing an example of a nuclear plant to which this improvement method is applied. Figure 2 illustrates a case in which preventive maintenance execution step S3, which is one of the equipment reliability processes in this improvement method, is performed. The execution details of preventive maintenance execution step S3 include hydrogen injection or a combination of hydrogen injection and precious metal injection into the BWR.

[0026] <An example of a configuration in which this improvement method is applied to a nuclear power plant> As shown in FIG. 2, a nuclear power plant P100 using a BWR includes a BWR reactor P1, a turbine P3, a condenser P4, 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. The reactor core P13 is loaded with a plurality of fuel assemblies (not shown). Each fuel assembly includes a plurality of fuel rods filled with a plurality of fuel pellets made from 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 located below the downcomer P17 and is connected to the lower space (lower plenum (not shown)) of the pressure vessel P12. The bottom of the lower plenum is called the reactor bottom. The structure that supports the shroud P15 is the shroud support P41.

[0027] The feedwater system is configured by installing a condensate pump P5, a condensate purifier P6, a feedwater pump P7, a low-pressure feedwater heater P8, and a high-pressure feedwater heater P9 in this order on a feedwater piping P10 connecting the condenser P4 and the pressure vessel P12. A hydrogen injection device P16 is connected to the feedwater piping P10 between the condensate purifier P6 and the feedwater pump P7 by a hydrogen injection piping P18. An on-off valve P19 is provided on the hydrogen injection piping P18. Furthermore, a precious metal injection device P31 is connected to the feedwater piping P10 between the high-pressure feedwater heater P9 and the pressure vessel P12 by a precious metal injection piping P32. An on-off valve P33 is provided on the precious metal injection piping P32.

[0028] The reactor cleanup system is composed of 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 purification device P27, installed in this order on cleanup system piping P20, which 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.

[0029] Reactor water present in the downcomer P17 inside the pressure vessel P12 is sucked into the jet pump P21 and led 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 the heat generated by the nuclear fission of the nuclear fuel material contained in the fuel rods of the fuel assemblies. Part of the heated reactor water becomes steam. This steam is led from the pressure vessel P12 through the main steam pipe P2 to the turbine P3, causing it to rotate. A generator (not shown) connected to the turbine P3 is rotated, generating electricity. The steam discharged from the turbine P3 is condensed into water in the condenser P4.

[0030] This water is supplied as feedwater through 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 purifier P6, the feedwater is further pressurized by the feedwater pump P7, and heated by the low-pressure feedwater heater P8 and the high-pressure feedwater heater P9. In addition, an extraction piping P14 is connected between the turbine P3 and the low-pressure feedwater heater P8 and the high-pressure feedwater heater P9. Extracted 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, respectively, via the extraction piping P14, and serves as a heating source for the feedwater. The extracted steam drain from the high-pressure feedwater heater P9 is returned to the low-pressure feedwater heater P8. The extracted steam drain from the low-pressure feedwater heater P8 is returned to the condensate pump P5.

[0031] The reactor water in the pressure vessel P12 contains metal corrosion products contained in the feedwater and products resulting from corrosion of the structural materials inside 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, which branches off from the recirculation system piping P30, driven by the purification system pump P24. These heat exchangers cool the reactor water to approximately 50°C. The cooled reactor water passes through the reactor water purification device P27, where the metal corrosion products contained in the reactor water are removed. After being heated in the regenerative heat exchanger P25, the cooled reactor water is combined with the feedwater flowing in the feedwater piping P10 and supplied to the pressure vessel P12.

[0032] When the operation of the reactor P1 is shut down, all control rods (not shown) are inserted into the core. By inserting all control rods, the nuclear fission reaction of the nuclear fuel material is stopped, and the operation of the reactor P1 is shut down. The heat remaining in the equipment inside the reactor core P13 and the pressure vessel P12 is removed by the evaporation of reactor water, but once the temperature drops to a certain level, the efficiency of heat removal by evaporation of reactor water decreases. Therefore, once the reactor water temperature drops to around 150°C, the equipment inside the reactor core P13 and the pressure vessel P12 is cooled using a residual heat removal system (not shown).

[0033] As described above, in the example shown in FIG. 2, the preventive maintenance execution step S3 in the equipment reliability process involves hydrogen injection or a combination of hydrogen injection and noble metal injection into the BWR (nuclear reactor P1). Hydrogen is injected into the feedwater from the hydrogen injector P16 through the hydrogen injection pipe P18 by opening the on-off valve P19. The hydrogen injected into the feedwater is sprayed into the reactor from a feedwater sparger (not shown) above the downcomer P17 and mixed with the reactor water. Because the gamma ray dose rate in the downcomer P17 area is moderate and promotes recombination reactions, the added hydrogen recombines with oxygen and hydrogen peroxide in the reactor water as it passes through the downcomer P17, generating water. In other words, the added hydrogen consumes dissolved oxygen and hydrogen peroxide, which are environmental factors that cause SCC, thereby mitigating the corrosive environment. As a result, the ECP of structural materials in contact with the reactor water is reduced, thereby suppressing the initiation and progression of SCC.

[0034] Precious metal injection is a technology that enhances the effects of hydrogen injection through catalytic action. For this, a solution containing a platinum (Pt) compound, for example, is injected from the precious metal injection device P31 through the precious metal injection pipe P32, opening the on-off valve P33, and then through the feedwater pipe P10 into the pressure vessel P12. The platinum injected into the feedwater merges with the reactor water, and some of the platinum particles adhere to the surfaces of structural materials that come into contact with the reactor water. The remainder adheres to the fuel surface. On the platinum particle surface, oxygen and hydrogen peroxide in the reactor water react electrochemically with the hydrogen to form water. When hydrogen is present in a stoichiometric ratio (H2O = 2:1) greater than oxygen (hydrogen peroxide is treated as half the amount of oxygen), the excess hydrogen becomes excessive, and the reaction between hydrogen and oxygen proceeds at a potential near the hydrogen oxidation-reduction potential (H2O) of -500 mV vs. SHE. Therefore, the ECP of the structural materials is mixed and falls to around -500 mV vs. SHE, achieving the target potential or lower in the preventive maintenance execution step S3. In other words, the ECP is brought into the target range. This suppresses SCC. SCC can also be suppressed by monitoring the ECP and performing the corrective step S4 to control the ECP so that it falls into the target range.

[0035] As shown in Figure 2, 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, corrosion potential sensors P35a and P35b are installed in the cleanup system piping P20 and the recirculation system piping P30 to clearly indicate, using ECP as a continuous monitoring indicator for the equipment reliability process, that the SCC suppression effect achieved by hydrogen injection and precious metal injection is being achieved.

[0036] 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. The bottom drain line P34 is used to draw reactor water from the bottom of the pressure vessel P12 and pass it through the purification system piping P20. In the nuclear power plant P100, a branch line P34a is provided from the bottom drain line P34, and a flange P36 is attached to the branch line P34a. An electrochemical corrosion potential sensor P35a is installed on the branch line P34a. 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 reactor water sampling line (not shown) or the like. Specifically, a flange P36 is installed on the recirculation pipe P30, and a corrosion potential sensor P35b is mounted on the flange P36 to measure the ECP of the reactor water flowing through the pipe.

[0037] Furthermore, in the nuclear power plant P100, electrochemical corrosion potential sensors P35c and P35d can also be installed in the lower plenum of the pressure vessel P12 and the core P13. The electrochemical corrosion potential sensors P35c and P35d are preferably installed in the neutron instrumentation tube P38. The electrochemical corrosion potential sensor P35c is installed at the position of the lower plenum inside the neutron instrumentation tube P38. The electrochemical corrosion potential sensor P35d is installed at the position of the core P13 inside the neutron instrumentation tube P38. Then, ECP measurements are performed by drilling holes in the neutron instrumentation tube P38 so that reactor water in the target region flows into these electrochemical corrosion potential sensors P35c and P35d.

[0038] The structural components of the reactor P1 are made of austenitic stainless steel, such as SUS316L. In addition to SUS316L, other materials used for the structural components of the reactor P1 include SUS304, an austenitic stainless steel, SUS304L, which is a SUS304 with a reduced carbon content, and SUS316NG (NG: for nuclear power plants), which is SUS316L with added nitrogen to increase its strength. Various electrodes can be constructed using these materials. Furthermore, for example, the shroud P15 is made of SUS304L or SUS316L. The cleanup system piping P20 and the recirculation system piping P30 are made of SUS304 or SUS316NG. Nickel-based alloys can also be used for the structural components of the reactor P1. The reactor bottom requires high strength, so Alloy 600 is used. Similarly, Alloy 182 and Alloy 82 are used as weld metals for the welds and overlays in the reactor bottom.

[0039] <<Specific details of this improvement method (equipment reliability process for SCC)>> Next, returning to Figure 1, we will explain the specific details of this improvement method (the equipment reliability process for SCC). The specific details of this improvement method will be explained using an example in which it was applied to the nuclear power plant P100, which uses a BWR. This improvement method can also be applied to pressurized water reactors and other nuclear reactors by simply changing the ECP continuous monitoring location to suit the reactor design conditions. For example, when installing a sensor in the ECP continuous monitoring location of a pressurized water reactor, if a sensor is installed inside the reactor, a corrosion potential sensor can be installed using a thermocouple tube from the top of the reactor to the top of the core, or it can be installed in the core. Furthermore, a corrosion potential sensor can be installed in the sampling line of the steam generator installed in the pressure vessel of a pressurized water reactor.

[0040] The component reliability process for SCC using this improvement method can begin at any step shown in Figure 1. For example, it is recommended to start with selection step S1 (classification of components or components). Selection step S1 involves selecting components or components with a high risk of SCC. In this selection step S1, the SSC of the nuclear power plant P100 preselects components or components with a high risk of SCC based on criteria related to safety systems and pressure boundaries. The term "safety systems" refers to the equipment that automatically (or manually) shuts down the reactor by activating the reactor emergency shutdown system when an abnormal condition, such as an overpower state or an abnormal increase in power, is detected during reactor operation, and that activates engineered safety facilities to protect the reactor core and containment vessel boundary in the event of an accident. The term "pressure boundary" refers to the primary reactor system, i.e., the pressure vessel and main cooling pipe system through which high-temperature and high-pressure coolant circulates.

[0041] The selection (risk assessment) in the selection step S1 can be performed using various risk analysis tools or by expert judgment. This allows high-risk components or parts to be prioritized and targeted in the equipment reliability process. For example, in a BWR, components 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. These components are considered high-risk when considering the materials used, stress, and water quality. Similar components that do not form a pressure boundary include the jet pump P21 diffuser, the jet pump P21 riser, the shroud support P41, the neutron instrumentation tube guide tube, and the weld lines of 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. Lower-risk SSCs include areas that are relatively easy to access and work on, such as the upper part of the reactor. In the selection step S1, the risk of these SSCs is classified, a priority is assigned for improving reliability, and the areas to be monitored are selected. The selection results (data) from the selection step S1 are sent to the monitoring step S2 and the improvement step S5. The monitoring step S2 and the improvement step S5, which receive the selection results, perform the processes described below.

[0042] In this improvement method, the selection step S1 is followed by the monitoring step S2. In the monitoring step S2, the ECP of the high-risk equipment or parts selected in the selection step S1 is monitored (measured) as a continuous monitoring indicator, and the crack growth rate is estimated from the results (performance monitoring, crack evaluation). Here, the continuously monitored ECP is used together with the material, stress, and crack length measured by inspection to evaluate the SCC crack growth rate. Since the ECP can observe the corrosive environment, it can be used to evaluate and estimate the crack growth rate of the monitored parts. The crack growth rate can be estimated, for example, using the crack growth rate diagram shown in the Japan Society of Mechanical Engineers' "Standards for Nuclear Power Plant Equipment, Maintenance Standards" or the Ford-Andresen equation.

[0043] This improvement method considers the "ideal performance" of a BWR with regard to SCC to be a state in which SCC does not affect BWR operation, and defines continuous monitoring to demonstrate that the ECP is in a state suitable for SCC suppression as an essential condition for improving equipment reliability. Material corrosion resistance and stress also affect SCC, but these cannot be continuously monitored during reactor operation. Instead, they must be measured during reactor shutdown. Therefore, they are not used as monitoring indicators in this embodiment. However, the measured values ​​of these material corrosion resistance and stress are stored in a database for evaluating crack growth rate and can be used, along with the actual ECP measurements, to evaluate crack growth rate. Furthermore, the reactor water electrical conductivity due to impurity ions also affects crack growth rate, so measured values ​​can also be used. Reactor water electrical conductivity is already continuously monitored by the reactor water's main equipment. Furthermore, crack-related inspection results are also stored in a database and used to calibrate the estimated crack growth rate. After monitoring is performed in monitoring step S2, the actual measured value of ECP and / or the evaluation results of the crack growth rate are sent to preventive maintenance execution step S3, corrective step S4, improvement step S5 and life cycle management step S6, and are referenced in each process.

[0044] In this improvement method, the monitoring step S2 is followed by a preventive maintenance execution step S3. In the preventive maintenance execution step S3, preventive maintenance is performed to reduce the ECP of equipment or parts with a high crack growth rate estimated in the monitoring step S2. In the preventive maintenance execution step S3, if a high crack growth rate is determined to be high in high-risk equipment or parts based on the constantly monitored ECP, the ECP is reduced. The reduction of ECP in the preventive maintenance execution step S3 is achieved by implementing corrosive environment mitigation techniques. Examples of corrosive environment mitigation techniques include the aforementioned hydrogen injection and precious metal injection. Other techniques, such as hydrazine injection and titanium oxide injection, have also been developed. By implementing these corrosive environment mitigation techniques, the ECP can be brought within a target range. An example of a target ECP range is -230 mV vs. SHE or less (see, for example, "Experience with Hydrogen Water Chemistry in Boiling Water Reactors," R.L. Cowan et al., Water Chemistry of Nuclear Reactor Systems 4, 1, p. 29, BNES (1986)). Furthermore, target ECP ranges, for example, for reducing the crack growth rate of stainless steel and nickel-based alloy weld metal 182 (182 alloy) to less than one-tenth of the rate without corrosive environment mitigation measures, can be set to -100 mV vs. SHE and -200 mV vs. SHE, respectively (Reference: "BWR Preventive Maintenance Guidelines (Method for Evaluating the Environmental Improvement Effects of Hydrogen Injection, 2nd Edition)," JANSI-VIP-18-2nd Edition, 2017, Japan Nuclear Safety Council). After preventive maintenance is performed in the preventive maintenance execution step S3, information about the execution is sent to the monitoring step S2. Upon receiving this information, the monitoring step S2 measures the ECP of the relevant equipment or part, evaluates the crack growth rate, and sends the measured ECP value and / or the crack growth rate evaluation result to the corrective step S4 and the lifecycle management step S6 to monitor whether the preventive maintenance execution step S3 is being effective.Upon receiving the actual measured value of ECP and / or the evaluation result of the crack growth rate, the correction step S4 and the life cycle management step S6 each perform the processing described below.

[0045] In this improvement method, it is preferable to perform a corrective step S4 after the preventive maintenance execution step S3. In the corrective step S4, the water quality parameters monitored in the monitoring step S2 are corrected so that the ECP falls within a target range. It is preferable to perform the corrective step S4 via the monitoring step S2 after performing preventive maintenance and implementing corrosive environment mitigation techniques in the preventive maintenance execution step S3. In the correction step S4, water quality parameters are corrected, such as by increasing the amount of hydrogen in the feedwater or the amount of precious metal attached, so that the ECP (and / or estimated crack growth rate) measured in the monitoring step S2 falls within the target range. The increase in the amount of hydrogen in the feedwater is achieved by hydrogen injection, as described above. In the case of hydrogen injection, the effectiveness of hydrogen injection changes as the distribution and intensity of radiation within the reactor changes due to core operation management. Therefore, the amount of hydrogen in the feedwater is increased to lower the ECP. However, there are cases where the amount of hydrogen in the feedwater cannot be increased due to restrictions on radiation dose. In addition, in the case of precious metal injection, hydrogen is added to the reactor water, as in hydrogen injection, so it may be affected by core management. Furthermore, in the case of precious metal injection, the effectiveness may be reduced due to peeling, detachment, or elution of the precious metal (mainly platinum) attached to the material surface. Therefore, in the correction step S4, the amount of hydrogen in the feedwater is increased and / or the amount of precious metal attached to the material surface is increased, thereby correcting the ECP so that it falls within the target range. In addition, in the correction step S4, hydrazine injection or titanium oxide injection may also be performed. In the hydrazine injection, the amount of hydrazine injected is increased to correct the ECP so that it falls within the target range. In the titanium oxide injection, the amount of titanium oxide deposited is increased to correct the ECP so that it falls within the target range. By correcting the water quality parameters in the corrective step S4, good equipment reliability can be maintained. After the corrective step S4 is performed, the monitoring step S2 checks whether the correction was performed appropriately, and if it is effective, monitoring continues. Information about the measures (corrective actions) taken in the corrective step S4 is sent to the improvement step S5 and the lifecycle management step S6. Specifically, upon receiving the information about the corrective actions, the monitoring step S2 measures the ECP of the relevant equipment or part, evaluates the crack growth rate, and sends the measured ECP value and / or the evaluation result of the crack growth rate to the improvement step S5 and the lifecycle management step S6 to confirm the effectiveness of the corrective action. The improvement step S5 and the lifecycle management step S6, upon receiving the information about the corrective actions and the measured ECP value and / or the evaluation result of the crack growth rate, respectively, perform the processes described below.

[0046] In this improvement method, it is preferable to perform an improvement step S5 after the correction step S4. In the improvement step S5, if the ECP does not fall within the target range even after the correction step S4 is performed, at least one of material replacement and stress improvement is performed to continuously improve the reliability of the equipment or parts selected as high-risk in the selection step S1. This makes it possible to reduce the SCC risk of the nuclear reactor as a whole. For example, in components or areas with extremely high neutron absorbed doses, the crack growth rate may not decrease sufficiently even if the ECP is reduced. Furthermore, there are areas where corrosive environment mitigation techniques are ineffective. For example, hydrogen injection or precious metal injection causes hydrogen to escape into steam when reactor water boils in the core, making it impossible to reduce the ECP in the area between the core and the feedwater mixing point where hydrogen is re-added. In such areas, material replacement or stress improvement is preferable through the improvement step S5. For example, material replacement involves replacing materials with more corrosion-resistant materials. Examples of material replacement include replacing nickel-based alloy weld metal 182 with improved nickel-based alloy weld metal 182 and replacing 304 stainless steel with low-carbon 316 stainless steel. Other methods include replacing piping and equipment, or weld overlays. Stress improvement involves applying compressive stress to the surface through peening, such as water jet peening or shot peening, or surface polishing or mechanical stress improvement (MSIP). Applying strong compressive stress to the material surface through peening can prevent SCC for decades. After the peening process, the initial inspection period begins, equivalent to that of a newly constructed plant. Once the improvement step S5 has been performed, information indicating that the improvement step S5 has been performed is sent to the monitoring step S2, the preventive maintenance execution step S3, and the life cycle management step S6. Upon receiving the information, the monitoring step S2 measures the ECP of the relevant equipment or part, evaluates the crack growth rate, and sends the measured ECP value and / or the evaluation result of the crack growth rate to the life cycle management step S6 to confirm the effectiveness of the improvement step S5. The life cycle management step S6, which receives the information and the measured ECP value and / or the evaluation result of the crack growth rate, performs the processing described below. Furthermore, the preventive maintenance execution step S3, which has received information indicating that the improvement step S5 has been performed, can also perform the preventive maintenance execution step S3 again following the improvement step S5. In this improvement method, the improvement step S5 can also be performed after receiving the selection result from the selection step S1.

[0047] In this improvement method, a life cycle management step S6 is preferably performed after the improvement step S5. The life cycle management step S6 monitors the ECP of the equipment or parts identified as high risk in the selection step S1 while repeatedly performing the selection step S1, monitoring step S2, preventive maintenance execution step S3, corrective step S4, and improvement step S5, thereby managing the crack growth rate. In other words, the life cycle management step S6 improves equipment reliability by linking each step in the above-mentioned equipment reliability process and continuously repeating the entire process over a long period of time. This reduces the overall SSC risk of the nuclear power plant P100 to an ALARP level. Furthermore, the life cycle management step S6 extends the time interval between SCC status checks by SCC inspections performed for preventive maintenance based on an SCC database created based on ECP data, inspection data, and SCC occurrence and growth data from other plants accumulated through the long-term implementation of the equipment reliability process consisting of the above-mentioned steps. This enables SCC crack growth rate management by replacing the time interval with ECP monitoring. Furthermore, in the life cycle management step S6, for equipment or parts that are constantly monitored, and for which inspections have revealed that cracks have progressed as assessed or that no cracks have been detected, the inspection intervals can be extended, assuming constant ECP monitoring. This improvement method allows for simultaneous improvement and rationalization of equipment reliability. This also makes it possible to review regular inspection items. This is a feature brought about by the equipment reliability process in this improvement method.

[0048] <<Example of equipment reliability process using this improvement method>> Next, a specific example of an equipment reliability process in this improvement method will be described with reference to Fig. 3. Fig. 3 is an explanatory diagram showing an example of the effect (graph of risk reduction) obtained by this improvement method over time.

[0049] As shown in Figure 3I, the selection step S1 first ranks the SSC SCC risk (classifying equipment or components). In the selection step S1, the risk rankings A to H for equipment, structures, or their components (hereinafter sometimes simply referred to as "components") are determined. In the selection step S1, the component A, which has the highest risk, is selected as the ECP monitoring index (corrosion potential monitoring component) in the monitoring step S2. It is recommended to select and designate components where ECP is unlikely to decrease. In this way, components prone to ECP decrease can be assumed to have an equivalent or greater decrease in ECP without directly measuring them. This reduces the number of installed corrosion potential sensors. Furthermore, the process is simplified because installation of corrosion potential sensors is not required in locations where installation is difficult. The determination of whether a component is prone to ECP decrease or not can be made appropriately based on past performance data. Furthermore, the determination of whether a region is resistant to or prone to a decrease in ECP may be made based on correlations between regions, analysis results using an analysis tool, etc. Any corrosion potential sensor can be used as long as it can measure the ECP in reactor water, etc. Then, in the monitoring step S2, the ECP of the high-risk area A selected in the selection step S1 is constantly monitored as a monitoring index, and the crack growth rate is estimated from the result.

[0050] Next, when corrosive environment mitigation techniques are applied in preventive maintenance execution step S3, the ECP of each part A–H is reduced according to the conditions of each part. Then, when the crack propagation rate is calculated based on the material, stress, and other conditions, and the risk is evaluated and estimated based on the decrease in ECP, the risk ranking changes, as shown in Figure 3, part II. Part A, which had the highest risk in Figure 3, part I, has its risk reduced to below the target (within the target range), as shown in Figure 3, part II, through constant ECP monitoring. If the risk assessed from the ECP deviates from the target range, specifically if the monitored ECP exceeds the upper limit of the target range, the water quality parameters are corrected in correction step S4 by increasing the amount of hydrogen injection or repeating precious metal injection to increase the amount of precious metal deposition, thereby reducing the ECP back to the target range.

[0051] Here, preventive maintenance execution step S3 and corrective step S4 may not reduce risk to the target range. For example, parts C and D shown in Figure 3, II, were selected as high-risk in selection step S1, and even after preventive maintenance execution step S3 and corrective step S4, the risk did not fall below the target. In this case, improvement step S5 is performed, as shown in Figure 3, III. In improvement step S5, at least one of material replacement and stress improvement is performed to reduce risk and continuously improve reliability. The fact that parts C and D do not fall below the target can be confirmed by measuring ECP, or can be determined based on previous performance or data from other plants. If the risk of parts C and D falls below the target through improvement step S5, the risk ranking will change again, and part E will become the part with the highest risk, as shown in Figure 3, III.

[0052] As shown in Figure 3, the lifecycle management step S6 repeatedly performs the selection step S1, the monitoring step S2, the preventive maintenance execution step S3, the corrective step S4, and the improvement step S5. The lifecycle management step S6 monitors the ECP of the part A identified as high risk in the selection step S1 and continues to manage the crack growth rate. In other words, in this improvement method, by continuously implementing these steps as an equipment reliability process, the risk of parts A through H becomes ALARP. Part A was the highest-risk part, but by implementing corrosive environment mitigation techniques and continuous ECP monitoring, the risk decreases along with other parts with lower risk than part A. Therefore, by continuously demonstrating the reduction in the risk of part A through continuous ECP monitoring, it is possible to demonstrate that the risk of other parts is also decreasing. In this way, in this improvement method, the reliability of the BWR plant is improved over the long term by implementing the above-mentioned equipment reliability process. Therefore, in this improvement method, the life cycle management step S6 constantly monitors the ECP based on a database of inspection results obtained in the equipment reliability process and other information, and on the premise of managing the crack growth rate, it is possible to promote rationalization while improving safety by extending the period until the next inspection of other parts, including part A, and reducing the amount of material to be inspected during regular inspections.

[0053] <Example of applying this improvement method to equipment reliability process in a BWR> Next, an example of applying the equipment reliability process of this improvement method to a BWR will be described with reference to Figures 1, 2, and 4. Figure 4 is an explanatory diagram showing an example of the effect (graph of risk reduction) obtained by applying this improvement method to a BWR over time.

[0054] First, as shown in Figures 1 and 4, I, selection step S1 selects components or parts with a high risk of SCC. In other words, selection step S1 ranks the SCC risk of SSCs (classifies components or parts). Specifically, as shown in Figure 4, I, the results of the crack growth rate evaluation determine the risk rankings for the components / structures or their parts as follows: the inner surface of the H7 weld line (inside H7) of the shroud P15 (Figure 2), the inner surface of the H4 weld line (inside H4) of the shroud P15, and parts A to F. Therefore, H7 is selected as the target for continuous ECP monitoring (continuous monitoring indicator) in monitoring step S2. Note that H7 is located in the lower plenum and is the part where the shroud support P41 and the shroud P15 are welded. H4 is the part where two cylindrical members constituting the intermediate body of the shroud P15 are welded. Therefore, in monitoring step S2, ECP is monitored using either the corrosion potential sensor P35c or the corrosion potential sensor P35a. Of these, the corrosion potential sensor P35c can measure the ECP at the location in the lower plenum that comes into direct contact with the reactor water, and therefore can measure the ECP more accurately. The corrosion potential sensor P35a is easy to install, but since changes in the water quality in the bottom drain line P34 (for example, decomposition of hydrogen peroxide, consumption of oxygen and hydrogen peroxide on the piping surface) can occur, it is preferable to handle the measurement values ​​taking this into consideration.

[0055] Next, as shown in Figures 1 and 4, II, in preventive maintenance execution step S3, precious metal injection is performed as preventive maintenance to reduce ECP. As a result, the ECP in H7, H4, and parts A through F is reduced according to the conditions of each part (II in Figure 4). When the crack growth rate is recalculated based on material, stress, and other conditions due to the reduced ECP, the ranking changes to H4, H7, and parts A through F, as shown in Figure 4, II. Here, continuous ECP monitoring clearly shows that the crack growth rate in H7, which had the highest risk in selection step S1, is below the target. If the risk (crack growth rate) evaluated and estimated from the ECP is not within the target range (specifically, if the monitored ECP is higher than the target), the amount of hydrogen injection is increased in corrective step S4, or precious metal injection is repeated to increase the amount of precious metal attached, thereby bringing the ECP back within the target range.

[0056] Because H4 is an area affected by radiation from the reactor core, crack growth rates are estimated using the formula for the irradiated material or the formula for when corrosive environment mitigation measures are not implemented. Therefore, it may not be possible to reduce the crack growth rate to the target range using corrosive environment mitigation techniques alone. For such areas, risk is reduced through improvement step S5, for example, by replacing the shroud with a structure without the H4 weld line or by using stress improvement techniques such as water jet peening (III in Figure 4). As shown in II in Figure 4, the crack growth rate in H4, which had the highest risk at this stage, will no longer develop if a shroud is replaced with a structure without the H4 weld line or compressive stress is applied through peening (i.e., the risk in H4 will be zero, as shown in III in Figure 4), and the risk will be reduced below the target through improvement step S5. This will change the risk ranking again, and both areas will be in the ALARP state. H7 was the area with the highest risk, but by implementing the injection of precious metals and conducting constant ECP monitoring, the risk has been reduced along with other areas with lower risk than area A, and so by continuing to demonstrate through constant ECP monitoring that the risk in area A has been reduced, it can be shown that the risk in other areas has also been reduced. Therefore, although not shown in Figure 4, life cycle management step S6 can promote rationalization while improving safety by extending the period until the next inspection of other areas, including H7, or by reducing the amount of material to be inspected during regular inspections, on the premise that constant ECP monitoring is conducted to manage the crack growth rate based on the database of inspection results and other information obtained in the equipment reliability process.

[0057] <<Variations>> In the monitoring step S2, at least one water quality parameter selected from the feedwater hydrogen concentration, reactor water hydrogen concentration, reactor water oxygen concentration, reactor water hydrogen peroxide concentration, reactor water electrical conductivity, main steam system dose rate, reactor water hydrogen / oxygen molar ratio, and reactor water excess hydrogen concentration, whose correlation with the continuously monitored ECP is periodically calculated in advance, can also be used as an auxiliary continuous monitoring index. The amount of material attached to the surface of materials to reduce ECP in the reactor can also be used as an auxiliary continuous monitoring index. This ensures equipment reliability in locations where ECP sensors cannot be installed or during periods when ECP sensors are temporarily unavailable due to malfunctions or other reasons. The relationship between ECP and the above auxiliary continuous monitoring indexes is affected by the power control status of the nuclear fuel installed in the reactor core, and changes depending on the nuclear fuel burnup, control rod position, core flow rate, etc., making it difficult to maintain the correlation over the long term. Therefore, it is preferable to periodically re-obtain these correlations at each fuel change, or at most every five years when all the fuel is replaced.

[0058] When hydrogen is injected in the preventive maintenance execution step S3, ECP is reduced by reducing the bulk oxidant (oxygen, hydrogen peroxide) concentration. On the other hand, when precious metals are injected in the preventive maintenance execution step S3, the precious metals attached to the surfaces of the structural materials act as electrochemical catalysts, electrochemically reacting oxygen and hydrogen peroxide with hydrogen in the reactor water to form water, thereby reducing ECP. Therefore, when precious metals are injected, it is recommended to confirm that there is a sufficient amount of attachment, and then select a water quality parameter indicating a stoichiometric excess of hydrogen over oxygen and hydrogen peroxide as an auxiliary indicator for ECP measurement. Therefore, it is preferable that the hydrogen / oxygen molar ratio and the excess hydrogen concentration (the hydrogen concentration equivalent to the hydrogen concentration minus the oxygen / hydrogen peroxide concentration in the stoichiometric ratio) are sufficiently hydrogen-excessive. When titanium oxide is injected in the preventive maintenance execution step S3, the titanium oxide attached to the surfaces of the structural materials reacts with Cherenkov light to reduce ECP.

[0059] Furthermore, the data obtained through the series of steps may reveal that the increase in crack growth rate is not due to ECP but to other water chemistry changes, such as electrical conductivity. In such cases, measures such as increasing the purification power of the reactor water may be implemented in one of the following steps: preventive maintenance execution step S3, improvement step S5, or corrective step S4.

[0060] Furthermore, as mentioned above, this improvement method can be applied not only to BWRs but also to ABWRs and the like. FIG. 2 shows an example of the application of this improvement method 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 cleanup system piping P20. As explained with reference to FIG. 2, in the case of the nuclear power plant P100 using a BWR, ECP is measured by providing a branch line P34a from the bottom drain line P34, installing a flange P36 thereon, and using an electrochemical corrosion potential sensor P35a mounted on the branch line P34a. In this regard, in the case of a nuclear power plant P100 using an ABWR, ECP can be measured as shown in Figure 5. Figure 5 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 Figure 5 differs from the nuclear power plant P100 shown in Figure 2 in the following ways: The nuclear power plant P100 shown in Figure 5 does not have a recirculation system piping P30 or a jet pump P21. The nuclear power plant P100 shown in Figure 5 has an internal pump P40 installed at the bottom of the pressure vessel P12. Furthermore, in the nuclear power plant P100 shown in Figure 5, the extracted steam drain from the low-pressure feedwater heater P8 is returned to the condensate pump P5. The extracted steam drain from the high-pressure feedwater heater P9 is returned to the feedwater piping P10.

[0061] As shown in FIG. 5, in the case of a nuclear power plant P100 using an ABWR, ECP is measured in the reactor water above the downcomer P17. To achieve this, in the case of the nuclear power plant P100 using an ABWR, a pipe P20a connected above the downcomer P17 and connected to the purification system piping P20 can be used. Specifically, as shown in FIG. 5, a flange P36 is installed on the pipe P20a, and an electrochemical corrosion potential sensor P35e is attached thereto to measure the ECP. Furthermore, although not shown, in the nuclear power plant P100 using an ABWR, a branch line can be provided on the pipe P20a, and a flange can be installed on the branch line to attach an electrochemical corrosion potential sensor thereto to measure the ECP. [Explanation of symbols]

[0062] S1 Selection Step S2 Monitoring Step S3 Preventive maintenance execution steps S4 Corrective steps S5 Improvement Steps S6 Lifecycle Management Steps P100 Nuclear Plant P1 reactor P2 main steam pipe P3 Turbine P4 Condenser P5 Condensate Pump P6 Condensate purification equipment P7 Water supply pump P8 Low pressure feedwater heater P9 High-pressure feedwater heater P10 Water supply piping P11 Reactor containment vessel P12 Pressure vessel P13 Core P14 Bleeding piping P15 Shroud P16 Hydrogen injection equipment P17 Downcomer P18 Hydrogen injection piping P19 On-off valve 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 P31 Precious metal injection equipment P32 Precious metal injection piping P33 On-off valve P34 Bottom drain line P36 flange P38 Neutron Instrumentation Tube P41 Shroud Support P34a branch line P35a~P35d Corrosion Potential Sensor P40 Internal Pump

Claims

1. a selection step of selecting equipment or parts that are at high risk of stress corrosion cracking; a monitoring step of constantly monitoring the corrosion potential of the selected high-risk equipment or portion as a monitoring index and estimating the crack growth rate from the result of the monitoring; a preventive maintenance execution step of executing preventive maintenance to reduce the corrosion potential for the equipment or portion where the estimated crack growth rate is high; a correcting step of correcting the water quality parameter so that the corrosion potential monitored in the monitoring step falls within a target range; an improvement step of continuously improving the reliability of the equipment or portion selected as being at high risk in the selection step by performing at least one of material replacement and stress improvement when the corrosion potential does not fall within the target range even after the correction step is performed; a life cycle management step of repeatedly performing the selection step, the monitoring step, the preventive maintenance execution step, the corrective step, and the improvement step, while monitoring the corrosion potential of equipment or parts that are newly selected as being at high risk in the repeatedly performed selection step, and managing the crack growth rate; and The life cycle management step extends the time interval between confirmations of the state of stress corrosion cracking by inspection based on a stress corrosion crack database constructed based on at least one of corrosion potential data, inspection data, and stress corrosion cracking occurrence and progression data in other plants, which are accumulated in the process of repeatedly performing the selection step, the monitoring step, the preventive maintenance execution step, the correction step, and the improvement step, and manages the crack growth rate of stress corrosion cracking by replacing the time interval with monitoring of the corrosion potential during that period. A method for improving the reliability of a nuclear power plant.

2. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the selection step comprises selecting in advance by at least one of a risk analysis tool and judgment by an expert based on at least one of a criterion related to a safety system and a criterion related to a pressure boundary for at least one of a structure, a system, and an equipment of the nuclear power plant.

3. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the monitoring step estimates the crack growth rate by using the corrosion potential, which is constantly monitored, together with material, stress, and an actual crack length measured by inspection.

4. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the monitoring step uses at least one water quality parameter selected from feedwater hydrogen concentration, reactor water hydrogen concentration, reactor water oxygen concentration, reactor water hydrogen peroxide concentration, reactor water electrical conductivity, main steam system dose rate, reactor water hydrogen / oxygen molar ratio, and reactor water excess hydrogen concentration as an auxiliary continuous monitoring index, the correlation of which with the continuously monitored corrosion potential being periodically determined in advance, and uses an amount of adhesion of a substance adhered to a material surface inside the reactor for the purpose of reducing the corrosion potential to supplement the continuous monitoring index.

5. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the preventive maintenance execution step includes performing at least one of hydrogen injection, noble metal injection, hydrazine injection, and titanium oxide injection in order to reduce the corrosion potential.

6. 2. The method for improving reliability of a nuclear power plant according to claim 1, wherein the corrective step performs at least one of increasing the amount of hydrogen in the feedwater and increasing the amount of precious metal deposition so that at least one of the corrosion potential monitored in the monitoring step and the estimated crack growth rate falls within the target range.

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