New Energy Power Generation Automation and Intelligent Technology

Taking the reactor dome flow monitoring instrument as an example, analysis and comparison is made on the routine lightning protection measures such that lightning protection devices are added in the design of signal acquisition circuit to improve the lightning protection availability of equipment and circuits and reduce the effect on DCS (digital control system). The practical application shows that this approach brings about minor effect on the signal acquisition circuit, features a flexible layout, convenient maintenance, and strong resistance to multiple lightning strikes, and is available for offering reference in the lightning protection of other equipment and apparatus.

The source range of the nuclear instrumentation system (RPN) is used to monitor the reactor neutron flux from shutdown to initial startup. When the count rate measured by source range detector is higher than the set value, a “high flux at shutdown alarm” will be triggered. This alarm alerts the operators that positive reactivity has been added to the core, such as unexpected dilution or control rods withdrawal. This paper analyzes the source range high flux at shutdown alarm setpoint update logic of a nuclear power plant. It is found that the setpoint will be updated to a low value after the source range is automatically put into operation, resulting in false triggering of the alarm. This paper discusses the avoidance measures of alarm false triggering from the perspective of operational administration and control logic optimization. The results show that both measures are feasible.

The instrument and control (I&C) emulation system that developed based on virtual technology is an effective mapping of physical I&C system. It can realistically reflect the properties of physical system, and can be applied to verify profit intelligent control strategies. In the stage of solution design, it is definitely necessary to analyze data storage and interaction procedures of each process or thread that represents various functions in system. Moreover, some simulation functions, for instance, saving and loading initial conditions, involve the management for various data types. Therefore, overall management for internal and external data is indispensable. Refer to above requirements, this paper proposes a set of data management methods that include interface design and process optimization for data interaction in I&C system. The method has been proven to be beneficial for data management in physical system, and it should have significant engineering application value in I&C system.

Core reactivity control and shutdown margin are provided by control rod clusters, which makes it a crucial system for reactor control and protection. The physical position of control rod is measured by rod position indication system (RPI). After equipment type change of rod control and rod position indication system (RGL) in a plant, an issue that fluctuation of measured position occurred even with a stable physical position, during the system commissioning. It will seriously interfere with the surveillance of control rod position. The root cause of measured position fluctuation is ascertained by measurement principle and circuit analysis, and a practical solution is proposed also. By acquisition of detector signals and development of analysis algorithm, the issue was solved. At the same time, optimization suggestions at the hardware level for the root cause are proposed.

The compatibility and anti-interference capability of wireless sensor networks within nuclear power plants in complex electromagnetic environments are the most essential elements to ensure reliable operation and effective combat of equipment within nuclear power plants in harsh environments. Interference signals generated by electrical, rotating and communication equipment and sensors attached to nuclear power plants, as well as malicious enemy interference sources, cause their electromagnetic environment to become increasingly complex. For the complex interference environment of nuclear power plant compartment, this paper firstly adopts supervised and unsupervised learning for interference cognition, and after obtaining the features of wireless interference sequences, anti-interference decision based on deep reinforcement learning and array antenna multi-domain anti-interference based on intelligent decision are implemented. From the intelligent interference simulation based on the physical transmission system, it can be seen that the proposed method can effectively identify and resist typical interference signals, which can bring reference to the anti-interference of nuclear power plants.

At present, the water levels of high-temperature and high-pressure vessels such as voltage stabilizers and steam generators on ships' nuclear power plants are measured by the pressure principle. Affected by the load conditions in the vessels, the differential pressure water level measurement has problems of low accuracy and low reliability. The radar type liquid level gauge has the advantages of high precision and high reliability, and is an effective means of measuring the liquid level of high temperature and high pressure containers. In this paper, a FMCW system split guided wave radar based on the principle of frequency domain reflection (FDR) is proposed for liquid level measurement. The antenna probe part and the electronics part adopt a split structure, and the electronics part is installed in the secondary instrument cabin with better environmental conditions., the radar is easy to meet the index requirements of shock, vibration and irradiation. The high-frequency signal emitted by the radar antenna is conducted along the coaxial rod probe, and the signal will be reflected only when it encounters the surface of the material, which effectively eliminates the influence of obstacles such as the heating plate in the container, the internal reference pipe water replenishment device and the reference pipe connection. Non-blind zone measurement of liquid level to meet the liquid level measurement requirements of high temperature and high pressure vessels in ship nuclear power plants.

As nuclear safety equipment for nuclear power plants, the drum screen can ensure the safety of the cold source of the nuclear power plant by arresting small debris in seawater. The drum screen is prone to failures such as corrosion of metal parts, large vibration of the driving device, and fracture of structural connecting bolts. In actual operation, due to the lack of automatic monitoring means, the operating personnel only rely on regular inspections to identify equipment operating failures, resulting in untimely detection of equipment failures, which will affect the reliability and safety of the cooling source of the unit.In this paper, by analyzing the operation failure types and root causes of the drum screen, the fault monitoring method of the drum screen is studied, and the vibration, stress, and corrosion monitoring system scheme of the drum screen is proposed; From the aspects of operation and maintenance, cost-effectiveness, etc., the feasibility of the condition monitoring of the drum screen is analyzed. This scheme can be used for safety monitoring and early warning of cold sources in nuclear power plants, realize the intelligentization of daily operation and maintenance of drum screens and reduce the operation and economic risks of nuclear power plants.

The drum screen is a coastal nuclear power plant commonly used water filtration device for the nuclear power plant cooling sea water, is one of the important equipment for cold source safety of the nuclear power plant. Due to the harsh operating environment and technical limitations, there are been a lack of systematic online monitoring methods on the status of drum screen. With the application and development of various new technology application in the nuclear power plant, it is possible to increase online fault monitoring of drum screen in order to meet the needs of information and intelligence on safety of nuclear power plant. This paper analyze the principles of different advanced measurement technologies such as the vibration monitoring, optical fiber/optical grating monitoring, the application adaptability and design of these various technologies on drum screen monitoring are discussed, and puts forward the main problems needing attention in the application of online monitoring of drum screen in nuclear power plant.

The source range channel of Nuclear Instrumentation System (RPN) of a nuclear power plant provides a neutron flux measurement at shutdown and during initialization phase of the start-up, which is very important to the control and safety of the plant. In order to qualitatively analyze the source range channel disturbance of the CPR1000 nuclear power plant, the article is based on the signal transmission link structure of source range channel, combined with the actual cases on the site, analyzes the easy failure points in the link to the source range one by one. Analyze the influx of the measurement channel, search the cause of the source range channel problem, and formulate corresponding treatment measures. It can not only analyze the safety and stability of the nuclear instrumentation system, and provide relevant information to the operator, but also has important guiding for the subsequent system commissioning, operation and maintenance of RPN system of similar types of nuclear power plants.

Each Instrumentation and Control (I&C) system shall be classified according to its suitability to implement I&C functions up to a defined category. At present, there is no uniform requirement for the classification of the Main Process Variables Control System (MVCS).The requirements of International Atomic Energy Agency (IAEA) and International Electrotechnical Commission (IEC) are summarized in this paper. The rationality of main process variables control function categorization and system classification is analyzed by safety analysis, reliability claim analysis and Probabilistic Safety Assessment (PSA) analysis. The analysis results show the reasonable main process variables control function categorization and system classification.

With the continuous development of nuclear power industry, various production data of nuclear power plant, such as reactor temperature, pressure and other data, also grow rapidly. All kinds of data are stored in the database in a structured form. This lays a data foundation for the application of big data technology in the field of nuclear power. “History is always strikingly similar”. From the perspective of time series, through the comparative analysis of the historical data of temperature, pressure and other indicators, we find that there are certain similar morphological characteristics, that is, the data go out of the similar development law in two different time periods. Therefore, through the mining matching algorithm, the transient data series in the current time period are compared with the historical retention time series to identify the most similar series in history, which plays a certain early warning role for the future. This paper intends to build an artificial neural network to find the best similarity calculation algorithm through training the massive historical label data and sample data, and finally achieve the effect of intelligent classification of production data, so as to predict the development trend in the future. The experimental results show that it is feasible to build a neural network model to realize the method of production data classification.

Vulnerability analysis is one of the important processes of cyber security risk assessment. With the gradual application of industrial Internet, intelligent technology, intelligent systems and equipment in the field of nuclear power, the methods of cyber attacks on nuclear power plants are gradually diversified, and the cyber security threats faced by industrial control systems are also increasing, and their cyber security vulnerabilities need to be continuously assessed. In this paper, combined with the current threat characteristics of nuclear power plant industrial control system operating environment, the cyber security vulnerability of the system is studied, and the typical vulnerability analysis items of nuclear power plant industrial control system are proposed to analyze the vulnerability characteristics of the system under different cyber security threat factors. At the same time, a vulnerability evaluation method based on functional safety impact analysis is formed, which takes the security level and consequence degree of the attacked system as the basic score, and the impact degree on the design of the defense-in-depth defense line as the weight for comprehensive scoring. Assess the vulnerability level to provide reference for the cybersecurity risk assessment and design improvement of the industrial control system of the nuclear power plant.

In this paper the circuit of output current of nuclear safety classification pressure transmitter was analyzed in detail, the mathematical model of pressure transmitter was built, and the 4–20 mA output current is linear function of capacitance sensor current. The calibration process was analyzed, a non-linear duality equation group in which the two positions of potentiometers are the variables was built, and in fact the calibration process is solving the equation group. Analysis indicated that the calibration process is an iterative process, the convergence of the mathematical model was analyzed deeply, the result indicated that for proper resistance network the convergence is good, and the fact circuit meet the convergence condition. For general calibration process, a optimization method of proper excessive zero current was provided to accelerate the process, analysis and test verified that the efficiency is promoted significantly.

The modeling of the secondary-loop system is important to investigate its behavior. Accurate modeling and simulation under transient conditions are also necessary to verify the control system of the secondary-loop and provide basis for operation scheme. Based on the 3KEYMASTER simulation platform, a simulation model of the secondary-loop system of a pressurized water reactor was built. The model includes the main subsystems such as turbine system, reheat heating system, condensate system, feedwater system. The control system with PID controllers is also added to the model. Three steady-state working conditions are obtained at 100%, 70%, and 50% full power. Power transients, such as power step change, are simulated by changing the boundary conditions, and the system responses are recorded and investigated. The trends of key parameters are reasonable and meets the simulation experimental requirements. The secondary-loop system model follows the basic physical laws and can be used for nuclear power plant engineering simulation analysis. With reasonable simplifications and assumptions, it can be used as a control object for nuclear power plant control scheme design and verification studies. Its steady-state accuracy meets the requirements of training operators and also provides a good test environment for the design and operation of new generation nuclear power units.

Steam generator level is the key and difficult point of nuclear power plant control. Improving the automation level of Steam generator level control can reduce the workload of nuclear power operators, reduce human errors and improve the work efficiency of operators. In the start-up and shutdown stage and low power stage of a GEN-II nuclear power plant, Steam generator is supplied with water by ASG system and operators manually operated control SG level. In a GEN-III nuclear power plant, Steam generator is supplied with water by AAD system in the start-up/shutdown stage, and SG level is automatically controlled by ARE Start-up control valve, thus it has realized the automatic control of SG level under all working conditions of the nuclear power plants and improved the intelligent level for the nuclear power plants. In the start-up and shutdown stage and low power stage of the nuclear power plants, the steam flow and feed water flow are extremely small, the measurement is inaccurate, and the conventional cascade control of SG level and feed water flow cannot be adopted. This paper mainly introduces the principle, test validation and optimization, problems and solutions of SG level automatic control in the start-up and shutdown stage of the GEN-III power plants.

This paper firstly analyses the characteristics of fire in the technical galleries and gutters of nuclear power plant, and based on the current situation of some nuclear power plant of China, it introduces the design features of technical galleries and gutters equipment layout, fire prevention zoning, fire detection, fire extinguishing facilities and smoke control and exhaust system. It then analyses the problems and insufficiency of these aspects of the fire protection system of technical galleries and gutters of nuclear power plant, and gives the relevant improvement suggestions.

Aiming at the incompleteness and inadequacy of the response time analysis method of the current nuclear power plant safety-level digital-control system (DCS), this paper proposes an analysis method based on probability theory for the response time of nuclear power plant safety-level DCS. Firstly, according to the structure and characteristics of the safety-level DCS, the probability model of the response time of the safety-level DCS is established, and the calculation method of the probability distribution of the response time is obtained. Then, according to the obtained calculation method and the parameters of the engineering example, the rang of the response time and the probability density function are calculated, and the response time of the engineering example is simulated, and the simulated response time rang, frequency and probability density function fitting curve are obtained. Finally, the response time of the engineering example is tested by the response time automatic test device, a large amount of data with statistical significance is obtained. After the data is processed, the range, frequency and probability density fitting curve of the response time are obtained. By comparing the theoretical calculation results, simulation results and test results of engineering examples. The correctness and feasibility of the analysis method are verified.

With the rapid development of nuclear power and the accelerated progress of decommissioning and treatment of corresponding nuclear facilities, it is inevitable to build a large number of nuclear fuel cycle facilities, so as to complete the production of nuclear fuel elements required by nuclear power plants and the recovery and treatment of spent fuel and corresponding three wastes (wastewater, waste gas and solid waste) generated by nuclear power plants. According to the nuclear power medium and long term development plan (2005–2020), China’s nuclear power industry follows the path of closed nuclear fuel cycle [1]. Based on this, this paper makes a preliminary analysis and demonstration of the rationality of the reactor closed nuclear fuel cycle.

With a consideration of alleviating the unstable control responses in traditional pressurizer control, this work adopts the cutting-edge deep learning method to optimize the PID control performance. A Long Short-Term Memory (LSTM) model is trained by data from a traditional PID control simulation and is then used to provide predictions to the PID controller such that the newly constructed intelligent controller can produce control signals for real time working conditions. The verification experiments conducted for both functionality and complex inputs successfully proved the advantages of the intelligent controller, showing up to 80.7% overshoot reduction and up to 60.73 s decrease in the control time for a steady state. Although its performance varies in different control cases, it does provide a deep learning-based control option for the pressurizer control in nuclear power plants.

Gateway is a protective equipment at the boundary of the nuclear safety-level I&C system. Aiming at the unknown threat across the boundary of information physical space, this paper studies several key technologies of active defense, and constructs the gateway deep secure architecture that can resist multilevel and multi-dimensional complex attacks. Firstly, the systematic control mechanism of the gateway is constructed, which mechanism includes access control for data flow, IP whitelist at network layer, function code whitelist at application layer, IP/MAC binding and the isolation mechanism of chip’s external ports. Secondly, based on the domestic trust root, the trusted chain from bootloader to the upper-layer application is built to guarantee the secure startup of the embedded system, and the key application processes are measured periodically and verified dynamically during the running process. Based on the above, from building equipment embedded trusted system environment to its security functions, a set of active security defense mechanisms of the gateway equipment sare formed, which enhances the security of the system and improves the information security protection n capability of the equipment.

Digital Instrument and Control System (I&C) is called the “Nerve Center” of nuclear power plant (NPP), which plays an extremely important role in ensuring the safe and reliable operation of nuclear power plant units. Therefore, the optimal control strategy and design scheme must be adopted for I&C in the design stage of combined heat and power retrofit projects in nuclear power plants. This paper introduces I&C design process and related technical evaluation for cogeneration retrofit of AP1000 NPP Units. It provides the key points of the renovation design of the I&C and also gives some suggestions. Because this project is the first I&C retrofit for combined heat and power among domestic nuclear power plants, it has great reference significance for the subsequent similar projects aiming at supplying district heat by nuclear power plants.

This paper briefly introduces the frequent occurrence of electrical fires in China, analyzes in detail five failure modes in electrical fires, such as short circuit, poor heat dissipation, poor contact, overload and leakage, and analyzes the current main domestic electrical fire monitoring methods according to the failure modes. By studying the serious problem of false alarm in the process of residual current monitoring, the current follow-up technology is creatively developed, which effectively avoids the “false alarm” problem of electrical fire monitoring caused by the change of inherent leakage current.

The shielding effectiveness (SE) of structural components is one of the important indicators of electromagnetic compatibility (EMC) performance, and it is related to the size, material, frequency and test points position in the structure. The measurement of SE is governed by a variety of factors. SE of structural components is affected by the measuring antenna size which is adopted to different frequencies and the testability of certain observation points. To address this issue, this paper introduces an example of the SE of DCS cabinet of nuclear power plant control system, where the neural network algorithm was used to build the relationship between spatial location and SE in the cabinet at specified frequency, so that the SE of unknown points could be predicted by the SE of known points. Besids, the model parameters were optimized by training with a large amount of data, the HPR1000 safety-class DCS cabinet was adopted for the verification, and high prediction accuracy was obtained.

The DCS configuration design process involves a large number of variables and multiple lists, and designers need to spend more time on the preparation, checking and maintenance of the lists. Manual compilation of files is a lot of work, and prone to human error. In this process, the use of automatic software to assist in the design can effectively improve efficiency. Based on the practical requirements of safety-class DCS design project of a nuclear power plant unit 1 and 2, this paper designs and implements a set of automatic generating and checking system of accounting list, which implements the automatic generation and typesetting of list files, and obtains good application effect.

Nuclear safety-class DCS realizes reactor safety protection function, which is directly related to the safe and stable operation of nuclear power plants, so its reliability should be analyzed and optimized. In this paper, the engineered safety features actuation system in nuclear safety-class DCS is analyzed through the control instruction logic flow involved in the structure, according to the components of the series and parallel or voting logic and other logical relations to bulid a fault tree model. Using fault tree model, the reliability degradation trend is obtained, and the minimum cut set, importance degree and other parameters are quantitatively calculated. According to overhaul cycle of nuclear power plant, a maintenance optimization method for engineered safety features actuation system is proposed, and some optimization suggestions are given.

Network security protection design for instrument and control (I&C) system in nuclear power plant is an important measure to ensure safe and stable operation of instrument and control system in nuclear power plant. Relevant national departments have put forward a number of regulatory requirements for network security design of instrument and control system, including the general principles of safety protection of power monitoring system proposed by Energy Bureau, Guidelines for grading of classified protection of cyber security by Ministry of Public Security and the technical policy requirements of network security of nuclear power plant proposed by China Nuclear and Radiation Safety Center. How to coordinate network security protection measures to meet multiple regulatory requirements is an important issue in the design of nuclear power plant network security protection. This paper introduces a forward design process of network security that coordinates and considers the above regulatory requirements, and analyzes and explains the design elements and reference standard for the design process, such as zoning and domain design, grading of classified protection, and critical digital asset classification protection.

The main helium fan of a high-temperature gas-cooled reactor uses a magnetic bearing as a shaft bearing, whose pulse signal of rotation speed could be easily disturbed by a high-power variable-frequency drive. There are also problems such as strong electromagnetic interference, potential safety hazards, and different potential distribution between strong and weak electric circuits. Moreover, analog signals require step-down to match digital signals when using AD conversion to transmit signals. This study uses optocoupler isolation to match voltage and improve the transmission quality of rotation speed signal without changing the circuit. Tests and verification have been done on the test maglev blower. The results show that under the same conditions, compared with no isolation, the application of optocoupler isolation is reliable and effective. The application of this technique provides a reference for voltage regularization and enhances the anti-interference ability of analog signal transmission in a complex electromagnetic environment, like maglev blowers.

The digital I&C system of nuclear power plants integrates a number of technologies such as computer, communication, control and display, optimizes the monitoring and operation of nuclear power plants, provides more advanced control and management means, and is the brain and command center of nuclear power plants. Hardware equipment is the basic component of the digital I&C system of nuclear power plants, and its reliability level directly determines whether the system can operate safely, reliably and stably. Reliability analysis of the hardware is carried out, and the failure mode and failure mechanism of the hardware are studied to improve the nuclear power plant. The reliability of digital I&C system is very important. Combining the functional safety concept and safety integrity level requirements of IEC61508, this paper predicts the mean time between failures (MTBF), studies the failure mode and effects analysis (FMEA) of the hardware of the digital I&C system of nuclear power plants, and finds and eliminates the hardware failures of the digital I&C system of nuclear power plants. Weak links can reduce the effects to improve the reliability of the nuclear power plant digital I&C system hardware and provide an effective verification method for hardware reliability verification.

With the development of nuclear power digital instrument control technology, IP core, as an indispensable part of FPGA chips, is more and more involved in the realization of important functions of instrument control products. Due to the high reliability requirements in the nuclear power field, it is necessary to fully verify the logical functions of IP cores. Traditional verification methods cannot meet the requirements of complex IP core verification quality and efficiency due to the use of directional testing and poor reusability. This paper proposes a method based on UVM construction of IP core verification platform, through the analysis of IP core application scenarios, and uses UVM components to realize the hierarchy and reusability, and finally, random test excitation generated by components is injected into the tested IP through the verification module, and the output data is monitored to check whether it is consistent with the expected value. This method can quickly build a verification platform, and the random excitation generated by the platform can fully verify IP. The paper takes the PCIe X4 IP verification of domestic FPGA as an example to explain how to use this method to build a verification platform and execute test cases. By practice, this method shortens the verification time and improves the verification efficiency.

The trend features, which are named trend symptoms later, of the operation conditions of critical equipment in a nuclear power plant can reflect its operation conditions, especially the potential failures. The trend features can be acquired based on the related physical quantities, such as temperatures, pressures, and flows, measured by sensors. When judging the trend symptoms of the operation and equipment, it is often necessary to select a suitable inputting data length. This paper proposes a method to determine the inputting data length based on wavelet analysis. The equipment history data are filtered by wavelet analysis to remove the noise and burr, then the smooth data curve after filtering will be segmented by a time series segmentation method based on local maximum and minimum, and the analysis results are obtained by linear fitting of the data within each segment. Then, the shortest segment of the segmentation results is conservatively selected as the reference value for the inputting data length of the corresponding equipment. The effectiveness of the proposed method is demonstrated by testing with several pieces of equipment history data. The results show that the proposed method can improve real-time performance while ensuring the accuracy of determining the symptom of equipment operation data.

After the Fukushima nuclear accident, the design concept of Defence-in-Depth of the nuclear power plant has further developed, and the safety requirements for level setting of Defence-in-Depth and for independence of inter-level have further improved. In order to ensure current Defence-in-Depth analysis always meets the latest international nuclear safety requirements, this paper analyzes the Defence-in-Depth level of DCS of nuclear power plant based on Swiss Cheese Model, and make a comparison with the Defence-in-Depth analysis schemes before and after a domestic project improvement, and proposes a Defence-in-Depth level-setting method on the premise of meeting the regulatory standards, which provides a valuable reference for the continuous improvement of the digital I&C system design of third generation of nuclear power reactor.

The parametric design method is applicable to the series, generalization and standardization of nuclear power major equipment, that can helps to improve the modeling efficiency in the 3D model modeling stage. The Double Sheathed Liner Sumps in the nuclear power plant, it`s has the above structural characteristics. This paper proposes the method and technology of parametric design, to carry out parametric analysis and Research on the dimension and assembly relationship of Double Sheathed Liner Sumps. Through the development technology based on C software, the purpose of parametric design for the 3D model of Double Sheathed Liner Sumps is realized. The application of this system tool, it`s can improves the efficiency of 3D design of Double Sheathed Liner Sumps, the 3D model generated by this system, can meet the standardization requirements of the design criteria in the design process, and effectively reduce the manual errors. Therefore, the parametric design concept and research results can be applied to other similar nuclear power equipment.

The reliable operation of the digital I&C system of nuclear power plants is of great significance to nuclear safety, but there is no industry-recognized method for evaluating and improving software reliability. This paper proposes a method which is depend on software reliability growth model, to calculate software reliability. At first this paper analyzes the cause and mechanism of software failure,then demonstrates how to descript software reliability mathematically. The second part of this paper explains two way to reduce software failures: fault-tolerant design and error-avoidance design. Furthermore, the third part proposes a software reliability assessment and improvement method based on the reliability growth model, which is used for the digital I&C system of nuclear power plants.

The traditional nuclide identification algorithm based on the characteristic peak has high requirements for users’ specialization, many steps and slow recognition speed. In recent years, although the new methods such as artificial neural network and Bayesian theory based on full spectrum analysis have high recognition speed and low requirements for user specialization, they rely on a large number of training samples, and the accuracy of measurement results will be affected when the measurement conditions of training samples and measured samples deviate. To solve this problem, an algorithm of nuclide number identification based on singular value method is designed. The algorithm is to analyze the time series of the collected data, which is mainly divided into the steps of acquiring array signals, constructing incidence matrix, calculating and sorting eigenvalues. The experiments show that the nuclide recognition time is short and the recognition is accurate. It can provide fast and accurate information in case of abnormal nuclear radiation.

In the design of non-safety I&C system of nuclear power plants, different safety classification and power supply sequence design need to meet the characteristics of system network redundancy and self-healing. Therefore, engineers often use ring network configuration in system network design. When the ring network design or network wiring is wrong, it is easy to cause a network storm, resulting in the abnormal operation of the main control unit of the first-floor control network control station, and the mis-operation of the control unit on the field equipment, resulting in nuclear power safety accidents. This requires that the main control unit of the control station could resist the attack of network storm. This paper provides a method to prevent network storm based on link layer. The abnormal packets are filtered in the network protocol layer of the control unit, and the network storm threshold detection algorithm is used to monitor the network storm attack in time, and the network loopback protection control unit is set. Experiments show that this method can effectively resist the attack of network storm. When the main control unit is attacked by network storm, it can normally complete the task of logic unit, ensure that the field instrument and control equipment is in safe working mode, and the system can automatically return to normal state after the network storm stops. At present, this technology has been realized and applied in the digital I&C system of nuclear power plants, which has good adaptability and broad development space.

The operation space for grinding the end of reactor core is very narrow in heavy water reactor nuclear power plant. The traditional manual grinding method are low efficiency, high radiation dose and poor grinding effect. This paper designed a teleoperation robot system for grinding the end of reactor core. The robot operates in a linear Delta configuration at the slave end. The robot system integrates irradiation camera and surface detection sensor which can provide compact configuration, real-time monitoring, precise positioning and good overall results. This design greatly reduces radiation risk and provides a safe and reliable automation system solution for grinding tasks.

The battery supplies power to the safety-level equipment and instruments of the nuclear power plant in the event of a power outage in the entire plant. At present, many nuclear power plants in China have the problems of unmeasured connection resistance, lack of quarterly inspection and lack of operation test in the three links of battery installation and debugging, daily maintenance and regular test. This leads to the shortening of the life cycle of the battery due to the lack of effective measurement, maintenance and testing, and the premature replacement of the battery is a waste of resources and increases the cost of equipment.

The civil engineering design of nuclear power plants has a long cycle, involves many disciplines, and the design process is complex. The traditional design methods and technical means are backward, mainly two-dimensional design means. The design activities are lack of specifications and coordination. The design data are stored and managed in the form of drawings, tables, calculation instructions, etc., scattered on personal computers, and difficult to share and reuse. This paper has carried out the research on the key technologies of nuclear power civil engineering design. It has proposed to manage design tasks in the way of task map. It has realized the standardized management of design tasks and calculation activities. It has unified management and scheduling of nuclear power civil engineering design tools. It has opened the data connection with external systems. It has eliminated information islands. It has realized the structural management and intelligent application of design data. It has constructed the refined collaborative design of nuclear power civil engineering, It has strengthened business collaboration. It has controlled design process and accumulated design data among various disciplines, improved the efficiency of nuclear power civil engineering collaborative design. It has not only saved costs, but also laid a solid foundation for nuclear power projects to go abroad.

Fluoride salt cooled high temperature reactor (FHR) is a promising advanced fourth generation reactor. Self-stability of FHR focus on the ability to converge to an equilibrium point without control loop, which is crucial for the design of control system and operation strategy of FHR. In this paper, the self-stability of the FHR is analyzed based on its physical characteristic using shifted-ectropy approach. Through theoretical analysis, it is proved that the equilibriums of FHR dynamics are globally asymptotically stable.

In the nuclear safety level I&C (Instrument and Control) area, the code generator plays an important role in developing the application software. The main function of the code generator is to translate the Lustre program corresponding to the graphical control algorithm into C code. Therefore, the reliability and quality of the code generator will directly affect the safe and reliable operation of nuclear power plants. In order to ensure the correctness of Lustre code, a static semantic analyzer is designed and developed to find the errors in source Lustre code. Its function is to check the correctness of the input source Lustre code, including identifier-checking declarations, identifier-checking programs, type-checking expressions and equations, clock-checking expressions and equations. According to the characteristics of Lustre language, the static semantic analyzer is designed and developed, and its function and correctness are verified by a large number of test cases. In this paper, mainly introducing the design and implementation of the static semantic analyzer.

Tianwan Nuclear Power Station (TNPS) Units 7&8 with total capacity of 2 × 1200 MW, is located in Lian Yun Gang City, Jiang Su Province in China, which will apply the pressurized water reactor of VVER-1200 (AES-2006). China Techenergy Co., Ltd. (CTEC) is the supplier of the main instrumentation and control system (Main I&C) for Units 7&8 of Tianwan Nuclear Power Station.

This paper describes the development process of the computer-based procedure system (CBPs) and the requirements of regulation and standards. Based on the operating experience of the digital control system in CPR1000 nuclear power plants, it puts forward the user needs of the CBPs of the second generation nuclear power plants in service in China to improve the level of automation, the convenience of operation and maintenance management, and the friendliness of human-computer interface. The requirements and constraints for updating and modification of nuclear power plants in service are analyzed, including consideration of six aspects, such as software and hardware platform performance, operator operation habits, operation procedure architecture, power plant operation and maintenance plan, verification and training, and safety and economic benefits. Different types of CBPs is compared and analyzed, and the updating scheme of the CBPs of the nuclear power plant in service is proposed, which is divided into three stages. The first stage is to realize a well used computer-based procedure system. The second stage is to realize a computer-based procedure system with higher level of automation. The third stage is to realize a highly intelligent computer-based procedure system. This paper introduces the development and implementation process of the first level updating scheme of the CBPs of CPR1000 nuclear power plants. And it briefly describes the design principles and main technical points of the transformation scheme. The verification plan is proposed based on different platforms and different objects. The future development direction of the CBPs is analyzed. And personal prospects and solutions are proposed.

In recent years, nuclear power project management mode of underground pipeline applied gradually BIM to save costs and improve efficiency. GIS technology was applied to show geographical environment and spatial position relationship of underground pipeline. In the article, through in-depth study of BIM and GIS technical theory, application platform for underground pipeline by opening source technology was built. The design rules and model characteristics of underground pipeline were taken as the research objects, and the technical feasibility for BIM + GIS technology was verified. The research would provide basic guarantee for intelligent management of nuclear power’s underground pipeline.