This paper describes development of a method based on machine learning for break diameter estimation in PWR loss-ofcoolant accidents (LOCAs). This estimation uses the data from a safety parameter display system (SPDS) that are transferred from power stations to the Japanese nuclear regulatory body. Two features extracted from the reactor coolant pressure p were used for learning: the minimum pressure decreasing rate dPmin before p becomes less than 4 MPa after the occurrence of the LOCA, and the time τ_dPmin between the point when p equals the saturated vapor pressure and the point when dPmin was obtained. The programming language MATLAB was used to extract these features from the SPDS data and to learn them using a support vector machine (SVM) with the 2nd order polynomial kernel function. There were several factors that cause deviations in the feature extraction results such as break position in the reactor coolant piping and the one-minute sampling timing of the SPDS. After data were simulated by the severe accident code MAAP4 and one-minute sampled, the input break diameter D was learned by the SVM with these features extracted from the simulated data. The learning result D* showed a relative error of plus or minus 13% and standard deviation of 5.6% under the deviation by break position and sampling error. It was confirmed that these learning results were valid even if the time interval from the reactor shutdown to the occurrence of the LOCA was changed from 0 to 30 or 60 minutes. It was judged that the learning results can be put to practical use.

Manual ultrasonic testing (M-UT) and Automated UT (A-UT) has been applied to nuclear component welds in In-service Inspection (ISI). M-UT, in which only an ultrasonic waveform is displayed, can easily be performed using a relatively economical device, so it is widely used for general purposes as well. However, M-UT examiner can note down records information (such as the ultrasonic waveform and defect positions), when found waveform exceeding the threshold level, but it’s not possible to take all waveform records in passed weld regions. Therefore, the reliability of inspection results of M-UT is affected by the skill of examination personnel.

In contrast, A-UT has the function of automatically saving to the PC all data obtained during inspection, including the ultrasonic waveforms, probe positions. Therefore, A-UT can prove there is no flaw in an inspection object by the digital recorded data, but A-UT is more expensive and takes workload for preparing than M-UT.

We developed an absolute recordable manual ultrasonic testing system (ARMUT® system) to take easily and economically records, to increase reliability for ultrasonic testing. The ARMUT system is already deployed to nuclear power stations, and oil and gas plants.

The replacement from lithium hydroxide (LiOH) to potassium hydroxide (KOH) for the pH control of the primary water of pressurized water reactors (PWRs) is under consideration because of an unassured supply and high cost of enriched ⁷Li. In this study, the susceptibility of the stress corrosion cracking (SCC) initiation of a Ni-based alloy in the KOH chemistry was investigated. In addition to the comparison of KOH and LiOH, the influence of the dissolved hydrogen (DH) concentration on the SCC initiation susceptibility in the KOH chemistry was examined. In the experiment, the constant load tests were performed for Ni-based alloy X-750 at 360°C in simulated or modified primary water with LiOH or KOH. In the LiOH environment, the DH concentration was set at 30 cc/kg, a normal DH concentration in PWRs. In the KOH environment, the DH concentration was set at 5, 30, or 45 cc/kg. As a result, the SCC initiation time showed no significant difference between the KOH and LiOH conditions at the DH concentration of 30 cc/kg. Interestingly, the SCC initiation time was significantly extended under the condition of the low DH environment (5 cc/kg). This DH dependence in the KOH environment was consistent with the reported DH dependence in the LiOH environment. The experimental result in this study indicates that the initiation of SCC can be mitigated with a low concentration of DH in the KOH environment as in the LiOH environment.

This publication analyses the influence of the geometry of the crack front on the propagation of a defect on a CT12 specimen made of Tu52b steel. The work is carried out using a numerical analysis based on the application of the Gfr criterion. This method, in the field of fracture mechanics, is used to simulate the ductile propagation of cracks in a structure.

The finite element simulation is performed using SYSTUS software, where different models are compared to observe the influence of the modelling of a non-straight crack front on the propagation of the defect, namely a crack front geometry determined by post-processing of the results. These models are compared to a model with a supposedly straight crack front.

A brief description of the Gfr criterion, initially defined in S. Marie's thesis (Reference [1]), is presented. Then, the results obtained using the model with a non-straight crack front are presented and discussed. It is shown that the impact of mesh refinement is low on the overall behavior of the specimen. A discussion of the J-integral values obtained is also done. Finally, a comparison with a straight crack front and a 2D model is used to verify whether the global behavior of a CT specimen is well represented by a model with a non-straight crack front. The overall behavior of the specimen will also be compared to analytical values by applying ASTM E1820 standard

During the shutdown period of pressurized water reactor nuclear power units, the film-forming amine maintenance technology can effectively inhibit the shutdown corrosion of secondary circuit carbon steel pipelines. In this work, octadecylamine was taken as the research object, and the inner wall of deaerator drain pipe of PWR nuclear power plant were selected as typical service conditions. The maintenance process and adsorption mechanism of octadecylamine on A515-60 carbon steel were studied by using material interface characterization method and thermodynamic calculation software. The influence mechanism of film-forming concentration, duration, carbon steel surface state on octadecylamine adsorption behavior was discussed. The research results are significant for the popularization and application of film-forming amine maintenance technology during shutdown of PWR nuclear power plant.

The maintenance effect of octadecylamine on A515-60 carbon steel and pre-oxidized A515-60 carbon steel surface using one-time dosing was studied by scanning electron microscope, X-ray energy scattering spectrum, high temperature electrochemical measurement technology, weight loss and contact angle measurement. It was found that the best filmforming concentration of octadecylamine on A515-60 carbon steel surface was 20mg/kg, and the best film-forming time was 8h at 168°C.

The adsorption layers of octadecylamine on A515-60 carbon steel and pre-oxidized A515-60 carbon steel were studied by focused ion beam-transmission electron microscope. It was found that the thickness of organic films adsorbed on A515-60 carbon steel in 20mg/kg octadecylamine solution at 168°C was 14nm, and the corresponding adsorption layers were 25. The surfaces of A515-60 carbon steel had multi-layer physical adsorption.

A circumferential crack was detected during an in-service inspection of the pressurizer spray line piping of a pressurized water reactor (PWR). Destructive investigation revealed an inter-granular crack with a depth of 4.4 mm, which was propagated in the weld heat affected zone (HAZ) of a 316 stainless steel. The line is filled with hydrogenated PWR primary water, assuming oxygen stagnation is not a primary cause. The detailed initiation mechanism is still unclear, and the propagation was considered inter-granular stress corrosion cracking of extremely high hardened HAZ. The potential cause of the high hardness is the superimposition of high heat input and constrain condition of nozzle shape.

The result of crack growth simulations using EPRI MRP458 explained reasonably the cracking process. The calculated propagation of a tiny crack from 0.5 to 4.4 mm depth was 9 years, which is within the operation time of the Ohi-3 (i.e., 19.3 effective full power year).

Within the development of the technology related to nuclear power plant, more and more attention have been paid on the design efficiency and reasonability. However, at present, the layout design of equipment in a typical nuclear power plant has always been carried out relying heavily on engineering experience, which is limited in the searching for candidate schemes and time-consuming. Thus, we propose an automatic layout method to realize the effective arrangement of equipment in nuclear power plant, in which some engineering concerns can be taken into account.

In the method, the digital model of each equipment with socalled double-space geometry was established according to the list provided by the user. The digital model servers to describe the geometric feature of the equipment to be placed in nuclear power plant. The algorithms of positioning were proposed to determine the relative position of each equipment to be arranged. By building the map for the path from one equipment to another, a shortest path combination will be determined among all possible paths based on the double-source shortest path algorithm. By applying this algorithm, the path for the workers in the nuclear power plant, or for the transportation of equipment can be determined accordingly. The results demonstrate and prove the feasibility of the proposed procedure.

Grayloc connections are part of CANDU Primary Heat Transport System (PHTS) and connect the inlet and outlet feeders at the End Fittings (EF) to the Fuel Channels (FC). In each CANDU reactor, the number of Grayloc hubs is in the range of 760 to 960. Grayloc leak has been a lingering operational issue in past decades, causing unacceptable amount of forced or extended planned outages.

Following multiple Grayloc leaks at Pickering Nuclear Generation Station (PNGS) since 2019, extensive investigations [1, 2, 3] have been conducted to find root causes or contributing factors to the leaks. In this paper, major potential contributing factors are identified, reviewed and discussed.

In the end, no single/sole contributing factor is identified in causing recent PNGS Grayloc leaks. Several major factors or more precisely - their combined effects were recognized for the recent leaks: weak original Grayloc design; local Flow Accelerated Corrosion (FAC) near the hub/seal ring contact surface to initiate the potential breakthrough leak path via steam cutting; hubs with improper installed Grafoil tape; higher feeder load causing accelerated stress corrosion; the extent of local FAC effect increasing with service years. Impact ranking of potential contributing factors is also given in the paper.

Nuclear power plant status monitoring and equipment fault diagnosis are important guarantees for the safe operation of nuclear power plants. With the development of artificial intelligence technology, fault diagnosis of nuclear reactor equipment has entered the stage of intelligent diagnosis and prediction. The control rod drive mechanism (CRDM) is the actuator of the reactor control and nuclear safety protection system. In the normal state, CRDM realizes the stepping movement of the control rod through the alternating action of the three sets of coil currents. When an abnormality in the coil current is detected, it means that the CRDM movement fails, and in more serious cases, it may cause the control rod to accidentally drop. Therefore, in order to prevent the occurrence of control rod stuck, sliding rod and driving failure in the reactor and to ensure the safe and economic operation of the reactor, a necessary measure is to effectively monitor the operation status of the CRDM. In this paper, the overall architecture of an intelligent perception and adaptive control system for CRDM is studied. By analyzing the multi-condition and multi-mode coupling relationship in the time dimension and spatial correlation between the coil current and vibration signals, a condition discrimination and fault diagnosis model for CRDM based on hybrid network of CNN+BiGRU is established. Furthermore, a digital adaptive control model of control rod drive system is built based on the real-time simulation system dSPACE. Through functional verification, the system can perform real-time perception and identification of normal operation conditions, fault diagnosis and adaptive control based on the coil current and vibration signals. The recognition accuracy rate for various operation conditions reaches more than 99.35%, and the control accuracy of the sequential current is improved around ±4% of the rated value under various external environment changes, providing a supportive reference for the subsequent nuclear reactor intelligent equipment development.

The point kernel integration method shielding analysis code cosKERNEL is one of the important code in the COSINE code package, it mainly used in the shielding design and analysis of the main circuit equipment and auxiliary system equipment of nuclear power plants. In order to ensure that the code can actually be used for shielding modeling calculations in nuclear power plants, it needs to be verified. The theoretical model and solution method of point kernel integration are introduced in this paper. After that, application of cosKERNEL code is carried out by using measured data of operating nuclear power plant and ANSI/ANS, ESIS international benchmarks. Comparison results show that the calculation results of the cosKERNEL code are more conservative than the measured values, and are in good agreement with the benchmark code calculated values, which are within the benchmark numerical range. The self-developed cosKERNEL code has high calculation accuracy and reliable results.

In the event of a steam turbine shutdown, pressure in the steam path of the turbine suddenly decreases to a vacuum. This causes pooled drain in the heater to immediately boil and turn into Flash-back steam coming back to the turbine, the so-called Flash-back phenomenon. This phenomenon can cause serious damage to long blades.

To verify the structural reliability of the blades, model steam turbine tests were conducted. The size of the model turbine was 1/4.2 of the 1500rpm full scale actual steam turbine, and rotated at 6300rpm to keep the velocity triangles the same as that in the full scale turbine. In terms of mechanical features, the blade stress of the model turbine was the same as that of the full scale turbine. The test facility contained two sets of a hot water tank and a connecting line to the turbine. Quick-move gate valves in the lines were opened during low load and full speed steam turbine operation. This caused preheated water in the tank to boil and come back to the steam turbine, which simulated the Flash-back phenomena. The Flash-back mass flow rate and wetness of Flash-back flow were varied as test parameters to investigate the influence of Flash-back on blade stress with various Flash-back flow conditions. The test parameters were decided considering the situation that the turbine is shutdown at 100% load condition which is the most severe condition in actual steam turbine operations and was not verified in previously reported tests.

The vibration stress was obtained from strain gauges mounted on the suction side of each blade. The relationship between Flash-back conditions and vibration stress was investigated. As a result, it was confirmed that the vibration stress on the long blades caused by the Flash-back phenomena is sufficiently low compared to the strength of the blade material.

Process Data Reconciliation (PDR) is the only tool for the reconciliation of all technical processes and other scientific applications to achieve "Contradiction-free and quality-controlled data with the lowest possible uncertainty". The theory of the PDR method is based on the Gaussian correction principle. Several examples will be shown on how PDR can be used in a customer-oriented way at nuclear power plants. The use of reconciled values as basis for correction factors will be described. The way to achieve acceptance by plant staff and regulatory authority will be discussed. One of the most important objectives in a PDR project is how to visualize the results and communicate this information within the organization.

A very minor leak from the primary to the secondary side of the steam generator heat transfer tube may allow a very small amount of primary coolant to enter the secondary loop system during reactor operation, contaminating the secondary loop system. While radionuclides entering the gas phase are ultimately discharged directly into the turbine plant through the turbine leakage or into the nuclear auxiliary plant ventilation system through the condenser evacuation and then discharged from the chimney after monitoring, radionuclides entering the liquid phase in the second-loop system are purified through the steam generator drainage system and the condensate purification system. As a result, for the radiation protection design of nuclear auxiliary systems and turbine plants, the second-loop source term analysis is crucial. The Hua-Long Pressurized Reactor (HPR1000), a third-generation nuclear power plant, was built using experience gained from Chinese commercial nuclear power plant designs, construction, operation, and maintenance. Based on the generation, transfer, and removal of radioactive substances in the HPR1000 unit's second-loop system, the calculation model for the radioactive source term in the second-loop system is established in this paper. The results of γ source intensity analysis in the cooling water of the secondary loop under different operating conditions are given. The findings can be used to analyze dose change, optimize staff occupational exposure dosage exposure during operation, and evaluate radiation protection for the secondary loop discharge system under various operating situations.

In order to validate capabilities of ultrasonic testing (hereinafter, UT) from outside surfaces of reactor pressure vessel (RPV) (ex-vessel UT) to detect cracks and determine the cracks size, Hamaoka nuclear power station unit 1 (Hamaoka unit-1) owned by Chubu Electric Power Co., Inc., which is under decommissioning work was selected to be used as a validation plant for ex-vessel UT since there were a lot of cracks found on the nickel based alloy material welds of shroud support plate to RPV and of shroud support plate leg to RPV, and in the future the capabilities of the ex-vessel UT will be evaluated by comparing with direct measurement results of the crack samples taken out from the plant. Examination procedures of ex-vessel UT for those welds in Hamaoka unit-1 were established separately by two Japanese RPV suppliers and each company performed ex-vessel UT to the weld of shroud support plate and shroud support leg jointing to RPV by their own procedures. In the validation test, the results of crack sizing showed significant differences between two companies. However, the differences could become smaller by conforming relative UT parameters of two companies to each other. It was assumed that the reasons to cause the differences were due to differences of detectability on actual field work, beam orientation and angle, beam frequencies, automatic or encoded manual ultrasonic examinations（EMUT）of scanning and decision criteria of boundary between low alloy steel (RPV) and buildup welds. The reason for the discrepancy in the crack sizing results would be clarified in the evaluation by comparing with direct measurement results of the crack samples.

In nuclear power plants, high reliability nuclear-grade motors is required drive critical equipment. In order to avoid serious accidents on rotating machinery, accurate rolling bearing fault detection index plays an important role in detecting incipient faults. This paper proposed a new non-manual index which can accurately detect incipient bearing fault based on vibration analysis. The detection method consists of three stages. The first stage is obtaining the residuals between the measured data and the reconstructed data. For obtaining the residuals, Auto-Associate Kernel Regression (AAKR) is adopted to reconstruct the current state based on normal state vibration signals. In the second stage, Sequential Probability Ratio Test (SPRT) is adopted to determine the instantaneous state of the bearing. A new non-manual method, which is named as Regression Distance Index (RDI), is proposed as the third stage to accurately locate incipient bearing failures. The results of experimental verification on the bearing database of the Center for Intelligent Maintenance Systems (IMS), University of Cincinnati show the reliability, accuracy and versatility of the index. It shows following advantages: (a) the index can detect incipient bearing fault accurately; (b) the index is self-adaptive without manual setting; (c) the index is robust for inner race, outer race and roller elements fault detection, as the results will not impact by the quality of training data for not using empirical models.

Direct vessel injection (DVI) device connected to the reactor pressure vessel has been applied in most passive pressurized water reactors. During the emergency cooling injection process, concentrated boric acid solution is injected to the reactor through the DVI tube, and it mixed non-uniformly with the cooling water from the cold tube in descent section of reactor pressure vessel (RPV). The non-uniform mixing will affect the reactivity distribution of the core and bring core return to critical state. Therefore, it is of great significance to know accurately the boric acid concentration distribution during direct injection in pressure vessel and to study the influence of the mixing process during injection in coolant to improve the safety of reactor operation. According to the structural characteristics of ACP1000 pressure vessel, the 1/6 scaled numerical model was established. The transient flow mixing process of direct injection solution and coolant in the annular descent section and lower chamber of RPV was obtained by numerical simulation method. Variation tendency of boric acid diffusion process changes with the Reynolds number of safety injection fluid and cold pipe fluid. Results show that flow rate ratio of injection fluid and cold pipe fluid mainly affect the mixing degree, and then dominated the concentration distribution of boric acid on the same section. Gravity causes the injected fluid to form an envelope in the vertical direction and gradually dilute and diffuse in the circumferential direction.

The test facility called experimental verification platform for prognostics and health management technology of nuclear power plants is based on the third generation PWR independently developed by China, and includes the main systems of four nuclear power plants, namely, Reactor Coolant System, Chemical and Volume Control System, Main Feedwater Flow Control System and Main steam system. The purpose is to provide a data set with high correlation, high complexity, high controllability and multiple operation modes for the development of predictive operation and maintenance technology. In terms of design, data that can fully reflect the characteristics of the complex system of the prototype power plant is obtained by keeping the system control logic and sensor configuration consistent with the prototype power plant. At the same time, the fault injection function at instrument level, equipment level and system level are realized under the premise of controllability, which fully makes up for the shortage of fault data in the prototype power plant. Using the data set generated by the test facility and the real operation data of the target system of the nuclear power plant, the verification of predictive maintenance technology can be more efficient, and the safety of on-site deployment can be improved.

At this stage, digital technology and intelligent technology have been deeply developed, and the automation level of platforms and equipment can widely support the automatic operation of nuclear power systems. At present, the control systems of thermal power plants and other industries have been greatly improved in automatic operation, while the automatic operation of nuclear power plants has not been widely improved in terms of operation and maintenance of nuclear power plants. It is necessary to increase the research on automatic operation of nuclear power plants. For the normal start/stop process of nuclear power plants, automatic start/stop is an important part of improving the level of automation, which can significantly improve user experience, reduce start/stop time, improve the availability and economy of nuclear power plants, and effectively avoid human errors. During the start-up process, the operation of increasing plant power from 2% to 25% FP power is a key operation in the start-up process of the power plant. From the perspective of general operation requirements, this paper carries out a task analysis for the start-up process of increasing power from 2% to 25% FP power. The general framework and general technical scheme of self-starting operation provides a research direction for the design scheme of HPR1000 automatic start-stop under all working conditions.

Life Cycle Management (LCM) is a decision-making method to optimize the service life of major equipment. Through the implementation of LCM, the state management of power plant equipment and the optimization of operation life are realized at the same time, and the economic benefits are maximized on the premise of ensuring the safety of equipment. One part of LCM is to evaluate the economy of various maintenance schemes, give the long-term economic indicators of each maintenance scheme in the future from the perspective of the financial budget of the power plant, and quantitatively analyze the possible cost of the power plant and the comparison of the required cost between each scheme. This part of work is the core of the whole LCM plan. The economic analysis of longterm operation and maintenance plan determines the best maintenance plan by calculating the cost of multiple maintenance plans including preventive maintenance cost and corrective maintenance cost. This paper introduces the economic analysis method of long-term operation and maintenance plan through the research on the content of LCM economic analysis published by EPRI in the United States, gives the calculation example and calculation results of Comanche peak power plant, and discusses the problems that should be paid attention to when the above method is applied to nuclear power plants in China, including failure rate calculation, refueling cycle and net present value. This method can provide a reference for the general method of economic analysis of establishing long-term operation and maintenance plan of nuclear power plants in China.

This study aims to achieve high-accuracy exergy distribution analysis of the NPP thermal system, providing a more reasonable reference for optimizing devices and conserving energy. Firstly, by using the first and second law of thermodynamics, the modular model of each component in NPP is established. To improve the accuracy, an elaborate model is built to calculate the reactor core's efficiency; the type of steam generator is innovatively in consideration (the natural recirculating steam generator and once-through steam generator); the turbine’s seal leakage which corresponds to section area of seal radial clearance, throttles number is in consideration; the extraction pipe’s steam parameters under off-design conditions are confirmed by the modified flugel formula. Also, for the pipeline, the equation of exergy loss which only needs the inlet pressure and temperature is derived. The overall model of the whole system can be obtained through a regular combination of these component models. Secondly, a simple and efficient model is established using C++ programming, the input parameters for each device are operating pressure, temperature, inlet, and outlet flow rate, and thermodynamic equilibrium quality. Then, a typical pressured water reactor is applied for simulation, the main parameters of NPP streamlines are determined and the exergy destruction coefficient and exergy efficiency of each component are obtained. This paper laid a foundation for establishing the general system exergy analysis model of pressured power reactors and the results show that the model is accurate and efficient.

The high temperature gas-cooled pebble-bed modular reactor (HTR-PM), considered as the first pebble-bed modular type demonstration nuclear power plant in the world with the safety characteristics of the fourth-generation nuclear energy system, has been constructed in Shidao Bay, Shandong Province, China. It connected to the grid first time on December 20, 2021. The inherent safety of HTR-PM largely depends on the adoption of the tristructural-isotropic (TRISO) coated particle as the nuclear fuel, in which the SiC layer can retain nearly all the fission products. However, there are still small amounts of nuclides which can diffuse from the intact or defective TRISO coated particles at a high temperature. Along with the radionuclides produced from the fission reaction of uranium contamination in the matrix graphite and neutron activation reaction in the core, they dominate the radioactivity in the primary coolant. The radiation monitoring on the radioactive level of the primary coolant is essential to assess the fuel performance in the reactor core. An on-line gross γ monitoring instrument was designed to monitor the real-time variations of radioactivity in the primary circuit, which can provide a direct reflect of the fuel element performance in the core. It adopted NaI detector, which was aligned with the center of the primary pipe. While, in the actual installation environment, the center of the pipeline to be measured was deviated from the detector with a maximum distance of 37 mm. Obviously, this deviation will decrease the detection efficiency of the instrument. We adopted the MCNP (Monte Carlo N Particle Transport Code) to calculate the detection efficiency under normal condition, and implemented a sensitivity analysis of the influence of the pipe deviation on the detection efficiency of the instrument. Current quantitative calculations can provide valuable guidance for the on-line radiation monitoring instrument installation in the nuclear power plant.

In safety system modifications of nuclear power plants by multiple companies, there is a concern of interference if they cannot recognize the latest status of on-site changing and the layout plan of other companies. Due to the lack of the understanding of other companies’ engineering, it will lead the on-site interference and require the re-work of engineering.

For reducing the risk of interference concern, we develop and provide the 3D space sharing Web application that can display 3DCAD models, point cloud, and 360° images from any view point. 360° images with shooting position information can be extracted easily by technology of estimating shooting position of 360° movies. The users can tag comments in its 3D space and can generate contact forms with URL link to the same view. These function enable collaboration to accomplish safety system modifications between multiple companies.

We provide the 3D space sharing Web application to electric power companies, cooperated companies and ourselves. The application supports remote engineers who cannot easy to go to site and cannot easy to hold face to face meeting by spread of COVID-19 to check on-site situation and other companies’ equipment layout plan and adjustment equipment layout plan.

The construction of high temperature gas cooled reactor pebble bed demonstration project (HTR-PM) in China began in 2012 and realized grid connected power generation in December 2021. Great progress has been made, which has accumulated experience for the promotion and application of subsequent high temperature gas cooled reactor technology. In particular, the operation performance of the first set of Innovative Nuclear Power Technology has been verified through the commissioning test on the project site, which has laid a foundation for the normal operation of subsequent high temperature reactor demonstration projects. As one of the main auxiliary process systems, the spent fuel storage system also adopts an innovative design scheme, and the rationality and operability of the system design are verified through commissioning tests. Among them, the residual heat removal ventilation system for spent fuel on-line unloading as a key technology has also carried out in-depth commissioning tests, The operation characteristics of the residual heat removal ventilation system in the unloading stage of the spent fuel storage canister under different working conditions are verified through the commissioning test, so as to ensure that the performance of the residual heat removal system in the unloading stage of the spent fuel meets the safety requirements, which is also of great significance to the safe operation of the whole high temperature gas cooled reactor demonstration project.

In the Onagawa Nuclear Power Station Units 1 and 3, it was confirmed that the spent fuel pool (hereinafter referred to as pool) water temperature converged to a certain temperature even when there is no cooling of the pool.

In addition, it was confirmed that the convergence temperature is well below the operating limit set by the safety regulations in the current fuel storage condition during long-term shutdown of the plant.

Furthermore, it was confirmed that the evaluation results from the newly constructed pool water temperature evaluation model were in good agreement with the measurement results.

As a result of the planned suspension of cooling of the spent fuel pool at Unit 1 of the Higashidori NPS, the water temperature was in good agreement with the evaluation results from the temperature evaluation model, and was stabilized at the temperature well below the limit value specified in the safety regulations. Based on the results, the impact on the plant when the residual heat removal system, which was on standby as a backup of the fuel pool cooling system during the long-term outage, is put out of standby, was evaluated, and it was confirmed that the impact would be small. Therefore, we reviewed the operation of the residual heat removal system and optimized the maintenance tasks. Through these activities, we are able to reduce the number of regular tasks and allocate management resources to more important tasks.

The suppression chamber and its internal structures, that is a vent system consisted of vent pipe, vent header, and downcomer at Onagawa Nuclear Power Plant Unit 2 of Tohoku Electric Power Co., Inc., were subjected to a significant increase in the acceleration response level applying to the seismic design due to the enforcement of the new regulatory standards in 2013. In order to ensure seismic resistance to the acceleration response level, a large-scale modification work of the components was required in addition to the upgrading of the evaluation method.

The main features of this modification work were followings. First, it was a large-scale modification work that our company or other companies had never done before. Second, the feasibility of the modification was carefully examined by using full-scale mock-up models. Third, the work plan was elaborated in consideration of the specific characteristics of the on-site work environment, and the work was conducted systematically under thorough management in preparation for the plant's restart.

Based on these efforts, the modification work is currently underway with the cooperation of the plant manufacturer (Toshiba Energy Systems & Solutions Corporation), who is in charge of the design and work, to complete the work as planned, while paying attention to safety work.

In recent years, titanium tubes of condensers have been damaged in many nuclear power plants in China. Heat transfer tube damage will cause condenser leakage, resulting in shutdown, power reduction or delayed startup events. It has a great influence on the normal and stable operation of the unit. The design margin of condenser heat transfer tube, the data of blocking pipe in previous overhaul and the tracking of defective pipe of some domestic units were investigated. The defects of condenser heat transfer tube and the influence of pipe blocking on equipment performance were analyzed. In order to ensure the safe operation of condenser, some suggestions on the management of condenser defect tube were put forward.

The conceptual design of a demonstration hydrogen production facility using heat supply from the high-temperature engineering test reactor (HTTR) is being researched and developed at Japan Atomic Energy Agency (JAEA). This facility produces hydrogen with a thermochemical water-splitting Iodine-Sulphur (IS) process that requires high-temperature heat to extract hydrogen efficiently. The HTTR could supply the heat to the IS demonstration plant through the secondary helium loop, coupling the IS plant to the HTTR.

The rated operation mode of the HTTR gives helium outlet temperature of 850οC with 660 effective full power days (EFPD). However, in order to achieve hydrogen with high efficiency, the helium outlet temperature should be as high as 950οC. Increasing the outlet temperature increases the reactor core temperature, and as a result the operation time decreases. If the operation time is reduced too much, it is not feasible to use the HTTR as a heat supply for the IS plant. Therefore, the purpose of this study is to estimate the operation time of the HTTR at high operation mode of 950οC to confirm whether it could supply long enough high-temperature heat for the demonstration hydrogen IS plant. The thermal-hydraulic model is also revised using the latest calculation method to improve the accuracy of the temperature distribution in the HTTR.

As results, the core temperature increase by about 50 to 100οC when the outlet temperature increases from 850 to 950οC. Although the increase of core temperature makes keff decrease by about 0.3 %Δk/k, the HTTR can still operate approximately 660 EFPD. Therefore, it is possible to use the HTTR for long-term high-temperature heat supply to the demonstration hydrogen production IS plant.

This research is carried out on the basis of the overall technical scheme for the life cycle management of HPR1000 Nuclear Power Plant. Aiming at the most important background evaluation and analysis link, the pipeline system is selected as the research object. Through analyzing the aging mechanism of the power plant pipeline system, it is determined that the thermal stratification phenomenon caused by the intersection of cold and hot fluids at the connection of the pipeline is the main reason for the thermal fatigue accident of the pipeline. According to Miner's linear cumulative damage theory, the method flow of component pipeline fatigue evaluation. Then, on this basis, this study selects the primary circuit T-tube as examples, uses the commercial software ANSYS Workbench to carry out numerical simulation analysis, establishes the model, loads the temperature load on the fluid solid coupling finite element model based on elastic thermodynamics, finite element theory, and fluid solid coupling theory, and considers the fluid flow and heat transfer of the pipe fluid, and the fluid solid coupling factors of the fluid and pipe wall interface heat transfer, The stress state of pipeline under different transient conditions is studied and analyzed, and the fatigue analysis and calculation are carried out to determine the fatigue cumulative use factor of components. The evaluation results have important guiding significance for optimizing the operation of nuclear power plant and improving the life management program system.

Since the passive nuclear power plant is not equipped with a boron recovery system, in order to minimize the amount of waste liquid generated by sampling, the inner diameter of the sample lines is usually small and the Grab Sample Panel is compact, which leads to the problem that the sample lines is easy to be blocked. The number of Containment Penetrations has greatly reduced to 2 for 11 liquid sample sources in passive nuclear power plant, which makes sample liquid vulnerable to the influence of other sample fluids, thus resulting in low sampling representativeness. In view of the problems, we have taken the following design optimization. First, a Spare Glove Box has been added to ensure the PSS sampling function when the MS20 is not available. The box is arranged in room 12152 of the auxiliary building, and two interfaces reserved to the radiological chemical laboratory were connected to it. It is connected in parallel and standby with the Grab Sample Panel. It can be used to obtain liquid and/or Containment atmosphere sample. Second, according to the characteristics of the sampling sources, the sample lines shall be reasonably separated, and the shared pipeline shall be reduced or avoided as far as possible to improve the representativeness of sampling. The liquid sample lines in the Containment joins into five sample lines, and three Containment Penetrations shall be added accordingly. Multiple isolation valves shall be set on the sample lines of a single sample point to avoid internal leakage and cross flow during the sampling process. Third, the other optimization measures also will be implemented. For example, the inner diameter of sample lines is increased, while the outer diameter of the sampling piping remains unchanged. The measure of added a Spare Glove Box has been successfully applied to the four operating units in China and have good results. The other measures have not been implemented, which will be applied to the subsequent new NPPs. The paper introduced the optimization measures to improve the convenience of PSS operation and maintenance and lower personnel radiation dose. It is a reference for other passive NPPs.

Background

When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without power. The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system.

Features of the Fusible Plug

The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements.

Future Actions

Hitachi-GE aim to introduce the fusible plug into domestic/overseas nuclear power plants in the future.

Construction work of Ohma NPP was suspended due to the Great East Japan Earthquake in March 2011 when progress rate of construction was 37.6%. At that time, due to the suspension of construction, some equipment was left unproper environment.

Though construction work was resumed in October 2012, construction areas were limited within the scope that would not be affected by the new regulatory requirements.

At present, we have no choice but to store the equipment for a long period of time at the construction site and the vendor’s factories. Under these circumstances, we’ve been addressing ageing management such as monitoring and trending of ageing effects and taking preventive actions to minimize and control ageing effects in cooperation with the venders. And appropriateness of the countermeasures was confirmed by unpacking curing and overhaul representative equipment.

The soundness of the equipment in storage will be finally confirmed through various tests along with construction progress in the future.

At Ohma NPP, it is assumed that current situation on product maintenance will continue for the time being. In Japan, there is no precedent for product storage during construction for such a long period of time, and we will continue to consider, implement, manage, and improve the measures, with the cooperation of the venders.

In our previous research, a new seismic fatigue evaluation method based on vibration velocity was developed. By this method, it can be shown that the effect of the high-frequency component of the input seismic motion on the fatigue damage could be small. Also, it was possible for the safe side to evaluate all the pipes of each plant by properly selecting the parameters. These previous results were reviewed in this report, and the following matters were reconsidered. It was suggested in the previous reports that the new method could be applied to general piping, but it was only confirmed analytically when a concentrated mass was attached to the tip of a cantilevered pipe, approximating a thin-walled pipe. In this paper, the new method could be analytically derived for more general straight pipes, and the effect of pipe thickness was investigated, and it was shown that the effect of pipe wall thickness could be small. The relationship between the ASME OM code, the new method, and the fatigue evaluation by CAV was considered on the fatigue diagram. Regarding the method of evaluating fatigue from the CAV of the input acceleration shown in previous studies, it was possible to give correct results if the input acceleration was sinusoidal, but the input seismic motion generally included broadband frequency components, and in this case, fairly conservative values were obtained.

Cold end optimization is an economic comparison and analysis of the cold end scale of different circulating water system configuration schemes based on design inputs such as specific site parameters, water intake and drainage conditions, environmental requirements, condenser heat load, economic operation period, and electricity price. Provide technical support for water system design. This paper takes one Nuclear Power Plant as an example to analyze the cold end optimization. Based on the introduction of some important equipment performance parameters optimized for the cold end, the main variables and calculation principles of the cold end optimization of the DC cooling circulating water system of the nuclear power plant are analyzed. The effects of different factors on the optimization results of the cold end were discussed, including the economics of optimization variables, the slightly increased power curve, the end difference, and the flow rate of the condenser tube, ineffective hypothermia etc., to provide a reference for the subsequent optimization calculation of the cold end of the DC cooling circulating water system of the nuclear power plant.

The demand for circulating water in nuclear power plants is large, and most of the domestic nuclear power plants built and under construction at present use seawater direct current (DC) water supply system. Circulating water pumps use more configurations of two pumps per machine, and all of them are unit system water supply method. If the circulating water system adopts the expanded unit system DC water supply, there is a large height difference between the condenser located in the turbine plant and the water level in the inlet pool of the circulating water pump room. The larger flow rate of circulating pump, long pressure pipeline and more valves are the other reasons. Therefore, in the process of various hydraulic transients may occur water hammer phenomenon and system water supply interruption and other accidents. In order to ensure the safe operation of the system, detailed water hammer calculation and transition process analysis are required, and suitable protective measures are selected. The aims of this paper is to (1) carry out transient calculations for the circulating water system of nuclear power plants with two pumps per machine to expand the unit system water supply; (2) analyze various conditions that may occur in the transient operation of the circulating water system; (3) carry out simulation calculations for various conditions; (4) use the output results as a basis for system control, equipment design, equipment procurement, and system operation programming. It can provide an important guarantee for the reliable and economic operation of the system and the nuclear power unit.

Three dimensional numerical simulations and comparative analysis are conducted to study the resistance and detailed flow field of the eccentric orifice plate. The results show that: (1) For single-stage eccentric orifice plate, the increase of eccentricity has less influence on resistance coefficient, but as eccentricity increases, the disturbance intensifies when the fluid flows through orifice plate, and the parameters such as pressure, velocity, and turbulent kinetic energy varies more drastically. (2) The resistance coefficient of multi-stage eccentric orifice plate is greater than that of the multi-stage concentric orifice plate with the same interspace. And the resistance coefficient of multi-stage orifice plate increase gradually with the increase of eccentricity. (3) With greater eccentricity, the resistance coefficient of multi-stage orifice plate decreases slightly until stabilizes while interspace increases. With less eccentricity, the resistance coefficient of multi-stage orifice increases gradually until stabilizes.

The function of the demineralized water production system of the nuclear power plant is to provide demineralized water meeting the water quality and quantity to the nuclear island, conventional island and BOP users. The relevant equipment of the demineralized water production system is arranged in the demineralized water production plant. Nuclear power plants are mainly built in coastal areas. Due to the increasing shortage of shoreline resources, it is increasingly difficult to select excellent plant sites. It is particularly important to efficiently use the area of existing plant sites. The optimal design of demineralized water production plants is of great significance for improving the overall resource utilization of nuclear power plants. Taking a nuclear power plant as an example, the design of demineralized water production plant is optimized by using a scheme shared by multiple units, and the analysis and comparison are made from the aspects of land occupation, cost, equipment layout, demineralized water production series settings, etc. The results show that the design scheme of demineralized water workshop shared by six units is the best.

In order to produce C-14 isotope, the first batch of commercial reactor C-14 irradiation production targets were put into the reactor and started large-scale production during the outage 211 of unit 2 of Qinshan 3rd nuclear power plant. In order to verify whether the newly replaced adjuster rods after C-14 target replacement meet the design requirements and ensure that the reactor with the new adjuster rods is operating in a safe state, a series of physical verification tests were organized and implemented, including the adjuster rods reactivity value measurement test and core flux distribution measurement test, etc. After the above tests, it is confirmed that the performance of the new adjuster rods is comparable to that of the original adjuster rods, and the reactor with the new adjuster rods is operating in a safe state.

The systems/equipment of nuclear power plant are complicated, the start-up and shutdown process is timeconsuming, the workload of the operating personnel is large, and human errors such as missed operations and malfunctions are prone to occur. Research and development of Automatic Nuclear Power Plant Start-up and Shutdown System (ANPS) can, on the one hand, reduce the workload of operation personnel and human errors such as mis-operation; on the other hand, it can standardize the operation procedures for start-up and shutdown of nuclear power plants and achieve "staff reduction and efficiency enhancement", which is of great practical significance to enhance the safety and economy of nuclear power plant.

This paper takes pressurized water reactor (PWR) nuclear power plant as the research object, and proposes a general technical solution of ANPS system for nuclear power plants based on the design experience of automatic start-stop technology in related industries and the specific needs of the nuclear power industry, including: functional requirements, usage requirements and functional architecture; then identifies common and dedicated key technical difficulties that may be involved in the development process of ANPS system, and conducts a preliminary assessment of technical feasibility.

The supporting cooling tower of a 1000MW nuclear power plant (NPP) in offshore is a super-large natural ventilation seawater cooling tower. The operation of the cooling tower will have an impact on the surrounding environment, and the salt deposition problem of seawater cooling tower is one of the current EIA concerns. In this paper, taking the super-large naturally ventilated wet cooling tower of a NPP as an example, according to the physical principle and formula of water vapor diffusion recommended by the American Nuclear Regulatory Commission, a mathematical model of the cooling tower was established to study the influence of salt deposition of the cooling tower by the key parameters of the cooling tower and the drop ratio (water eliminator efficiency), The results show that reducing the water drifting rate of the water collector （less than 7 per million） is an effective measure to reduce salt deposition.

Qinshan secend nuclear power plant developed a new pressurized water reactor (PWR) power distribution measurement technique to achieve 6-hour dynamic flux measurement. During this period, the Qinshan 2nd carried out detailed verification and analysis based on the measured data. The dynamic flux measurement method is proved to be safe and feasible, and can replace the traditional steady flux measurement method. In March 2022, dynamic flux measurement technology was applied for the first time in OT215 of unit 2 of Qinshan 2nd. The test results were qualified and the expected target was achieved. The new method reduces the waiting time of low-power platform in start-up stage and improves power generation capacity and economy of power plant.

In recent years, nuclear power plants are gradually transforming to intelligence, and the requirements for data transmission volume, real-time transmission distance and stability are further increased. As the next generation mobile communication technology, 5G can meet the performance requirements of large capacity, massive connection, high reliability and low delay in various scenarios. It can 'significantly improve the data transmission efficiency between any two points and reduce the data transmission cost between any two points'. With the gradual maturity of 5G technology, the application of 5G technology in nuclear power plants has become a trend. After having the idea of 5G technology and nuclear power integration, it is necessary to further demonstrate the application requirements of nuclear power for 5G technology. Based on the design characteristics and requirements of nuclear power plants, this paper studies the potential application scenarios of 5G in nuclear power plants.

Nuclear Power Plant abnormal diagnosis operation strategy (ADOS) is a general operating strategy used to diagnose the abnormal operation reason reliably and quickly in abnormal operation condition. By using PWR event oriented ADOS as research background, a decision tree program was developed to realize intelligence and automatically design. Furthermore, the intelligence design result was compared with the original ADOS, and the difference between them was analyzed. It is a beneficial attempt for machine learning technology application in nuclear power plant operation field.

As the important support system, the research on the power supply of the active PWR Nuclear Power plants is an important aspect to ensure that the design of the power plant can meet the response capacity of the design extending conditions, which has an important impact on the operation of safety measures, and is of great significance to prevent large-scale radioactive release of nuclear power plants. Loss of off-site power (LOOP) is considered in all plant design basis and managed through a series of redundant and diverse approaches. Station blackout (SBO) is a design extending condition (DEC) for the design of most nuclear power plants. The designed SBO response time, effective measures must be taken to make the unit safe shutdown state and prevent core melting. The minimum acceptance outage time of Station Blackout of the a third generation active and passive PWR Nuclear Power plants is studied in this paper, and also the response capacity is considered.

To further the effort of coal-to-nuclear (C2N) research, the pros and cons of integrating small modular, high temperature gas cooled, and molten salt nuclear reactor technologies with existing coal plant infrastructure at the Limestone coal plant near Jewett, Texas, USA, were analyzed. Reactor operating characteristics were compared against the Limestone coal plant. The technical requirements and feasibility of various integration schemes were assessed to determine what, if any, reactor technology is optimal for C2N at Limestone. In this analysis large light water reactors were excluded as the size and capacity of these reactors adds regulatory, licensing, and technical complexities that likely make most C2N integration schemes infeasible for these designs.

Results from this analysis imply that there is no one-size-fitsall approach to nuclear technology integration with Limestone. Each nuclear design being investigated has advantages and disadvantages that have to be weighed against cost, risk, and C2N integration complexity. Furthermore, the technical specifications of the coal plant in question, in addition to the requirements for maintaining or replacing existing plant components, will heavily influence which nuclear designs and transition schemes are most suitable for a given C2N project. Future work can build on this research by performing decommission analyses of specific coal plants to determine what components can be reused and which will need to be replaced, factoring in cost and risk of integrations, and performing a detailed plant integration analysis of the specific reactor technology at specific coal plants.

At present, the classification of nuclear power plant equipment mainly considers nuclear safety, and there is no equipment classification system aiming at economy. With the continuous development of the third generation of nuclear power plants are becoming higher and higher. As one of the most important economic indicators of a power plant, availability intuitively reflects the economic benefits of power plant operation. In this paper, a set of classification system based on availability is created, which classifies power plant equipment according to the influence of single equipment failure on the availability of power plant. By putting forward different technical management requirements for different equipment with different availability levels, the purpose of improving unit availability levels, the purpose of improving unit availability is achieved. This method is described from three aspects: identification of important equipment for nuclear power plant availability, classification of availability levels, and design management requirements for different availability levels. The equipment with important availability can be identified from the aspects of unplanned outage and planned outage through failure mode and effect analysis and critical path analysis of refueling overhaul. The classification of availability level is based on the identification of the important equipment of availability rate, which is divided into different levels according to the influence degree of equipment on availability rate, including immediate shutdown caused by equipment unavailability, power reduction or defense of unit in violation of operation technical specification, overhaul activity or extension of equipment maintenance time, reduction of redundancy and increase of shutdown risk. Finally, according to the different availability levels of equipment, relevant measures are taken to improve the availability of equipment from the aspects of design, manufacturing, operation and maintenance, quality assurance and so on, so as to achieve the effect of improving the availability and economy of the power plant. The equipment availability classification system is an equipment management system aiming at the economy of nuclear power plant, which covers the whole process from the initial design to operation management of the plant. The analytical approach adopted in this paper fills the gap in the hierarchical approach aimed at increasing availability. At the same time, the availability classification, as an economy-based equipment classification management system other than safety classification, is the basis of nuclear power plant performance monitoring, preventive maintenance implementation, corrective maintenance, equipment reliability improvement, and life/obsoleting management.

Rotating machinery is a large class of important equipment widely used in nuclear power plants (NPPs), and its reliability and stability are crucial to the operation of NPPs. To accurately identify the operation state of rotating machinery, a fault diagnosis method based on wavelet transform and deep residual neural network is proposed. Firstly, observation data is obtained through a plurality of vibration signal measurement sensors arranged at different positions of the rotating machinery, so as to analyze the operation state of the rotating machinery more comprehensively. Secondly, continuous wavelet transform is used to extract time-frequency features from multi-channel vibration signals, and the multi-channel time-frequency features are fused into time-frequency images. On this basis, the deep residual neural network is applied to adaptively extract the fault features contained in the timefrequency images to achieve accurate rotating machinery fault diagnosis. In this study, the motor experimental device and the rotor experimental device are used as test platforms to provide data support for the condition monitoring of rotating machinery.

The experimental results show that the proposed method can accurately identify the operation state of rotating machinery, and the effectiveness of the method is verified. The advantages of the multi-sensor monitoring strategy are further illustrated by comparison with single-sensor diagnostic effects. In addition, the method studied in this paper is compared with several current mainstream intelligent fault diagnosis methods. The comparison results show that the overall diagnosis effect of the proposed method is the best, which shows the potential application value of the method in the fault diagnosis of NPPs rotating machinery.

Aiming at the lack of data on centrifugal pumps in nuclear power plants and the inability to provide data support for the subsequent development of centrifugal pump status monitoring systems in nuclear power plants, this paper designs and builds a centrifugal pump test bench. On this basis, the impeller cavitation wear fault of the centrifugal pump is set up, and the acceleration signal of the normal state and the fault state of the centrifugal pump is collected by the acceleration sensor. Since the collected acceleration signal will be mixed with a certain amount of environmental noise, which will affect the monitoring effect, it is necessary to denoise the signal before performing feature extraction on the acceleration signal. This paper first aims at the difficulty of selecting the K value and α penalty factor of the Variational Mode Decomposition (VMD) algorithm, proposes to use the Whale Optimization Algorithm (WOA), and uses the envelope entropy as the fitness function to optimize the VMD model. On the basis of the above work, the collected acceleration signal is decomposed, the environmental noise is eliminated, and the time domain characteristic parameters are extracted from the recombined signal to complete the normalization process of the time domain characteristic parameters. Then the preprocessed time-domain characteristic parameters obtained by calculation are brought into the Support Vector Machines (SVM) model to complete the model training, and then complete the development of the nuclear power plant centrifugal pump condition monitoring system. The final experimental results show that the nuclear power plant centrifugal pump condition monitoring system developed in this paper has a high monitoring accuracy.

Inadvertent withdrawal of a control rod cluster was observed during lifting of the upper core structure in the 15th outage of pressurized water reactor (PWR) Ikata Unit3 on January 12, 2020. This event would be caused by magnetite sludge in the head of control rod cluster, which was brought from corrosion of drive shaft in control rod drive mechanism (CRDM). The sludge adhered to the head of control rod cluster and led to incomplete separation between the cluster and the drive shaft. To prevent the events due to the corrosion products, we carried out some experiments and their consideration to accumulate the corrosion characteristics for type 410 stainless steel (SUS410), which was the main material of the drive shaft, in the PWR primary coolant circumstances, whereas less knowledge for the corrosion was available at the present time. As a result, it was elucidated that SUS410 material was corroded quickly under the conditions with weak acid and saturated dissolved oxygen (DO) concentration at 473 K, which was very similar to the circumstance of the drive shaft inside at start-up stage of Ikata Unit3, providing the corrosion product of magnetite. Countermeasures to reduce the magnetite formation were also suggested.

The fast and accurate response of the temperature sensors important to nuclear power plant (NPP) safety can ensure that the control room make timely response in case of large process temperature transients. At present, the response time of temperature sensors in most nuclear power plants is measured during plant maintenance with plunge method under laboratory conditions, although it does not reflect the sensors’ real performance under operating conditions.

This paper studies the in situ measurement of response time of resistance temperature detectors (RTDs) based on noise analysis technology, which measures the response time with no interference to in-service RTDs. It can be used to identify degradation or fault of RTD response time without affecting the operation of the plant and to remind maintenance personnel to fix or replace the sensors in time, so as to ensure the safe and stable operation of NPPs. In this paper, the intensity and frequency distribution of RTD noise signal under different installation modes are analyzed, and the influence of acquisition frequency on power spectral density is studied. In this paper, the noise analysis method is used to analyze the output data of temperature sensor collected by nuclear power plant, and calculate and generate PSD curve, which provides convenience for obtaining response time. This method relies on measured data and processing algorithms, and is in line with the intelligence needs of future generation NPPs. It has potential benefits for the operation and maintenance of NPPs.