Nuclear Power Plants: Innovative Technologies for Instrumentation and Control Systems

In this paper, the two-way converged wireless communication network (WCN) schemes for three cases in nuclear power plants (NPPs) based on the combiner are investigated. Firstly, the applications of WCN in NPPs are listed, including conventional voice communication and other extension demands. Then the respective advantages and disadvantages of the existing mainstream wireless communication technologies in NPPs are described. Therefore, we need to retain the existing single wireless communication technology and converge other wireless communication technologies via combiners. Secondly, the paper discusses the converged WCN scheme for the cases where the coverage area is small without blind area. Thirdly, the paper describes the converged WCN scheme for the cases where the coverage area is large and feeder installation is unrestricted. Lastly, the paper discusses the converged WCN scheme for the cases where the coverage area is large, but multiple-feeder installation is restricted and signals can be only transmitted by a single feeder. The above three WCN schemes can converge the installed and other wireless communication technologies under different circumstances. Due to the complementation of different wireless communication technologies, various kinds of intelligent business requirements in NPPs can be satisfied and reform cost can also be reduced effectively. Furthermore, it is of great significance to enhance the reliability and performance of wireless communication system in NPPs.

This paper has introduced the life cycle of I&C system and related quality control measures to guarantee the reliability during development and validation phases, analyzed software and hardware reliability prediction models, discussed the reliability demonstration method applicable for critical safety system and put forward a reliability demonstration analysis for safety critical system in nuclear power plant.

In certain special scenarios, there will be a mixed radiation field formed by a variety of nuclides or rays. The monitoring system for such radiation fields requires the capacity of comprehensive and automatic analysis. In this paper, a specific mixed radiation field is given, which contains gamma ray, neutron ray, α aerosol and tritium in a high radon concentration environment. An intelligent monitoring system is developed to measure and evaluate the complex radiation field rapidly and accurately. The radiation performance of the system is verified by experiments, which achieves the expected index requirements.

With the wide application of nuclear technology, the radiation field shows more and more characteristics of hybrid, complexity and versatility. In order to ensure the safety and data transparency of radiation field, online radiation monitoring system is usually established for all important radiation fields. Its purpose is to monitor the parameters of radiation field in real-time and accurately, and evaluate the status of radiation safety scientifically and reliably. Therefore, a basic model of online monitoring system for mixed radiation field is established, which can be carried out proper expansion in accordance with the different scales and styles of radiation field.

The performance of power amplifiers for active magnetic bearings (AMB) is closely related with their main power circuits and control algorithms used. A three-phase bridge circuit is selected as a main power circuit of an AMB power amplifier. Through a space vector pulse width modulation (SVPWM) algorithm, two electromagnet coils can be driven simultaneously by the three-phase bridge circuit. However, when the coil current command signal for the amplifier fluctuates sharply, a specially designed duty ratio restriction strategy is required. In this paper, two duty ratio restriction strategies, which are Period Bisection Strategy and Proportional Reduction Strategy, are proposed for the situation that the coil current command signal exceeds the tracking range in one PWM period. Experiments show that when the change rate of the coil current command signal is large, the Proportional Reduction Strategy makes the distortion of the current in each coil (i.e. the difference between the required current change and the actual current change in one PWM period) proportional to the change rate of the command signal. And the Period Bisection Strategy significantly reduces or even eliminates the distortion of the current in the coil with a slower change rate of the command signal, but at the cost of increasing the distortion of the current in the coil with a faster change rate of the command signal.

The nuclear power plant simulator displays the thermal-hydraulic conditions of nuclear power plant in 2D. To enhance the 3D effect of the thermal-hydraulic display of nuclear power plant, the real-time simulation of thermal-hydraulic mechanism of nuclear power plant is studied based on 3D visualization. Firstly, a 3D model is established that can express thermal-hydraulic characteristics; secondly the real-time data is connected to the simulator through a program interface; finally Unity3d is used to integrate the 3D model with the received data in real time, and display it in a data-driven 3D model. At the same time, taking the primary circuit equipment of a nuclear power plant as an example, a 3D visual test is carried out on the thermal-hydraulic mechanism of temperature, pressure, and flow rate. The results show that this method can well display the thermal-hydraulic mechanism in 3D, providing a new idea about the training of nuclear power plant personnel.

As the nerve center of nuclear power plant, the quality of digital instrument and control system is directly related to the stable, safe and efficient operation of nuclear power plant. At the same time, with the advancement of the independent research and development process of instrument and control system, it puts forward higher requirements in terms of security and reliability. It is very important to evaluate the software quality of instrument and control system scientifically and reasonably. Therefore, a software quality evaluation model suitable for nuclear power instrument and control system is proposed in this paper. On the basis of three-tier architecture, by restricting the relative importance of the two major indicators of security and reliability, the weight vectors of the indicators of each tier are calculated through consistency test. Then the experts grade all the indicators of the lowest tier one by one. Finally weighted sum is made from bottom to top to calculate the software quality score of the instrument and control system. The method presented in this paper plays a guiding role in the software quality evaluation of instrumentation and control system in nuclear power plant.

The wireless communication technology has been widely used in daily life, bringing great convenience to our lives. Its low cost, easy maintenance and high flexibility can also bring benefits to the nuclear industry [1]. This paper analyzes the wireless network application prospects and benefits in nuclear power plants. For the application of wireless networks in nuclear power plants, the security risks of wireless networks are introduced, and the specific requirements of domestic and foreign network security standards on the application of wireless networks in industrial fields are combed. Finally, the security protection strategies and recommendations of wireless networks in nuclear power plants are proposed.

Cyberattack threats to nuclear power plants (NPPs) are becoming increasingly common and serious. Even if the digital instrumentation and control (I&C) systems have adopted preventive measures such as physical isolation, the risk of cyberattack is still unavoidable, which has significantly affected the safe and stable operation of the NPPs. Cyberattack identification of I&C systems is one of the most important steps in cyberattack assessment. As the network structure and system functions of the I&C system become more and more complex, the attack methods are also more and more diversified. The cyberattack identification by a single information source is difficult to identify various types of cyberattacks. It is necessary to integrate information from multiple sources through information fusion methods to obtain more comprehensive and reliable cyberattack identification. The paper uses the D-S evidence theory to integrate data from different sources, including both cyber and physical spaces. The Dempster’s rule of combination is used to synthesize the basic probability distribution of each evidence, and the attack type identification is also performed. Three typical types of cyberattacks, including denial of service, packet injection, and feedback spoofing, together with normal operation status are identified in this paper. To obtain the basic probability assignment (BPA) of each attribute, the triangular fuzzy number method is adopted. The experimental results show that the method adopted in this paper can effectively fuse the information from both cyber and physical spaces, and can accurately identify the four categories cyberattacks.

With the development of nuclear technology, the existing fissile nuclides can not meet the long-term needs of human beings. The development of fast neutron reactor using fissionable nuclides is the trend of nuclear energy development. Small modular Pb-Bi Fast Reactor (SMPBR) is a fast neutron reactor with liquid Lead Bismuth Eutectic (LBE) as coolant. In order to become a competitive reactor, SMPBR is usually characterized by high power density, strong load following ability and convenient transportation. Due to the current lack of relevant experience of lead-bismuth fast reactors in China, there are certain difficulties in the design of lead-bismuth fast reactors and the design of control systems. So in this work, the control system for the SMPBR is based on the sodium-cooled fast reactor. Firstly, the influence of control rod, inlet flowrate and inlet temperature is studied by single variable method respectively. Secondly, according to the relationship between the parameters, the control system of regulating power of moving control rods and not moving control rods is designed. Then, a robust PI controller is designed according to the dynamic characteristics of reactor. Combining the influence of control rods and flow on power and power to design controllers with different operating schemes.

In this work, the functional requirements of the integrated management of physical protection system has been studied, by analyzing the composition of the physical protection system, as well as the development status and trend of the integrated management system. The design scheme is studied from two aspects of the composition of the physical protection integrated management system and the software architecture, so as to provide reference for the construction of the integrated management system of nuclear facilities.

In the process of designing instrumentation and control cables, cable selection is usually carried out according to engineering experience, and the influence of cable laying length and core cross-sectional area on signal transmission is often ignored. In view of several typical terminal equipment in nuclear power plant, this paper summarizes the factors affecting cable laying length in engineering practice. By analyzing four typical signal control loops DI/DO/AI/AO, and special temperature measuring loops such as thermocouple and thermal resistance, the limitation requirements of wire resistance and cable maximum laying length in different control loops are calculated. The calculation method can be used as a reference for the selection and rationality analysis of the cross-sectional area of instrumentation and control cables.

With the rapid development of modern computer technology, control technology and communication technology, it has become a practical problem to optimize the monitoring mode of BOP sub-items, to seek more reasonable and effective, more economical and convenient, more stable and reliable engineering application solutions that reflect the mature automation technology today and to uniformly monitor and design these sub-items. In particular, it has been mature in thermal power plant applying auxiliary control network technology in auxiliary workshops, and the technology has been widely used in chemical industry, metallurgy and other industries. BOP sub-items are networked on the basis of the current monitoring mode, and without affecting the project progress. Realizing of moderate centralized control will improve the level of automatic operation and management of nuclear power plant, and meet the requirements of informatization management of the plant.

Human Factors Engineering (HFE) design of Main Control Room (MCR) is very important to the performance of operators and the operation of Nuclear Power Plants (NPPs). In order to reduce potential human errors, the design of an MCR must consider appropriate human factors. Human factors verification and validation (V&V) evaluations comprehensively determine that the HFE design conforms to HFE design principles. HFE V&V enables plant personnel to successfully perform their tasks to assure plant safety and operational goals. However, in the design verification process, the evaluation of some qualitative design indexes is largely limited by the subjective factors of the tester. In order to improve this deficiency, this paper presents an AHP-fuzzy complex evaluation method based on nuclear industry standard and MCR design experience to assist MCR HFE design verification. The proposed method provides a set of MCR HFE evaluation indexes, the determination of weight factors and fuzzy complex evaluation model. Some preliminary evaluation results are provided and discussed.

This paper explains the overall control strategy of nuclear island ventilation system in CPR1000 NPP (Nuclear Power Plant), and expounds the current situation of nuclear island ventilation system. It analyzes the problems to be solved and challenges to be faced. Based on the market products, this paper puts forward the new control strategy and idea. It is necessary to develop nuclear safety-class electrical heater power regulation control cabinet used for local temperature regulation control, and local control cabinet based on simple hardware or an integrated safety-class electrical heater with self-control. The safety-class local power regulation control cabinet and on-off local control cabinet were developed successfully in this study, but there are not products on the safety-class integrated electrical heater with self-control in market, which shall be developed in the future. Then, propose a control logic architecture, and introduce the control logic design idea how to implement from the equipment level to the plant level. At present, in new construction NPPs, nuclear island ventilation systems have basically realized the “system level” control level and completely realized the “sub-function level” control according to this study. Subsequently, the “system-level” and “plant-level” relationship can be sorted out to further realize “plant-level” control. More intelligent ventilation and air conditioning algorithms may be also introduced to improve the overall control level of nuclear island ventilation system in NPP.

Circulation water filter system (CFI) is a very important supporting system in nuclear power station, all the nuclear power station at home locates at the seaside so as to ensure sufficient cooling water supply. The system design of CFI in different unclear power station at home almost adopts the technology of M310 unit introduced from France. This document introduces and analyses the design discrepancy with CFI system functional requirements, and finally give the improvement solution in the third generation of nuclear power station.

As the nerve center of nuclear power plant, the control system ensures the safety and economy of power plant operation, and its software test is very important. With the development of digital nuclear power and the wide application of embedded system software in the world, more and more embedded software has been applied to the control system and intelligent equipment of nuclear power plant. This paper analyzes the characteristics of embedded system software of control system in nuclear power plant and the difference between embedded system software and general application software from the perspective of test, and the impact on test. It puts forward a unit test method of embedded software, which realizes the separation of application software and underlying hardware by means of software structure analysis, program piling, syntax modification, hardware simulation, etc. in nuclear power plant based on ARM architecture. The successful application of V&V project unit test of plant control system software provides reference for unit test of embedded and similar hybrid programming software.

The design of Instrumentation and Control (I&C) system in nuclear power plan needs to consider the problems of Common Cause Failure (CCF) and independence of itself in addition to meeting the overall Defence in Depth (DiD) target of the plant and ensuring the correct and reliable implementation of monitoring, control and protection functions under various operation conditions. The requirements of International Atomic Energy Agency (IAEA) are summarized in this paper. A design scheme of DiD for I&C system of Nuclear Power Plant (NPP) is introduced. in this paper. The compliance of DiD design for I&C system with the requirements of IAEA is analyzed. The analysis concludes that the I&C design scheme basically meets the requirements of IAEA. At the same time, the improvement of diversity should be further studied. The study of this paper provides valuable reference for the continuous improvement of the design of I&C systems in NPP.

Fukushima accident has caused serious damages to public health and environment. Licensees of nuclear power plants in China are required to take measures to prevent severe accident happening. Control and detection hydrogen in containment under severe accident is one of the improvement actions. This article has analyzed the requirements for hydrogen detection system and related tests requirements for this detection system. A comparison of IEC 61298 and IEC 61779 is made to find tests should be carried on a new developing detection system/equipment. Nuclear qualification tests also have been considered. One testing system frame has been put forward.

The primary loop system of nuclear power plant is a large-scale control system with serious coupling between loops and complex objects. In order to solve the problem of integrated control of primary loop core power, stabilizer pressure and water level, and steam generator water level, this paper applies improved predictive control algorithms to each sub-control system on the basis of mechanism modeling. For the strong non-linear problem of water level control, a multi-model predictive controller is constructed through linearization at each typical operating point. The simulation results show that the application of advanced predictive control algorithm to the primary loop control system has achieved good control performance, and it also lays a solid foundation for subsequent practical engineering applications.

For the problem of poor control effect of conventional PID controllers in complex systems, this paper fully combines the advantages of deep learning in feature extraction, regression prediction, and predictive control in dealing with multivariable, strong coupling and other issues. Firstly, the predictive model controller is constructed through the stacked sparse auto-encoder (SSAE) deep network, and then the stepped generalized predictive control (SGPC) is used as the rolling optimization controller. Thus, the SSAE-SGPC neural network predictive controller is built. Finally, a simulation experiment is carried out in the dual-input dual-output multi-variable coordinated control system of nuclear power plant. Through a series of simulations such as set value disturbance, internal disturbance and external disturbance, compared with conventional PID control, SSAE-GPC controller has better control effect.

System Availability, Rejection rate and Malfunction rate are vital technical indicators for safety class DCS when evaluating the reliability of the control system to meet technical requirements. This trend has attracted great attention of equipment manufacturers, regulatory agencies and users. Since there is no relevant standards or norms to guide the restriction at this stage, there is no unified verification idea in the industry. Although domestic and foreign equipment enterprises have tried relevant verification research, it still remains at the level of theoretical analysis and the confidence is not high. In the research of test method, if the traditional test simulation method is used for index verification, it often requires a lot of manpower, material resources and time. Therefore, this method is also regarded as inoperable, so it has not been able to form a reasonable and mature test of the verification scheme for a long time. Based on the general principle of reliability analysis, this paper makes a comparative analysis of the objects of reliability test, and proposes a reliability analysis verification and implementation scheme of safety class DCS.

In the process of human factor design and preliminary verification of the control room console in Nuclear Power Plant, traditional 2D human model design and analysis method is commonly used. However the results turn to be inaccurate and time-consuming. With the help of human simulation tool Jack, quick and precise analysis can be achieved. This paper puts forward the human factor engineering analysis of main control room console in nuclear power plant based on Jack including the visibility, accessibility, and collision detection. By comparing this method with traditional 2D design, conclusion can be drawn that Ergonomic Analysis of main control room console Based on Jack is more comprehensive and accurate, thus largely increasing the design efficiency and quality.

The third generation nuclear power reactor protection system (RPS) based on computer technology, that is, safety DCS system, is designed with large-scale integrated circuits and real-time embedded software, which has the characteristics of complex structure, large system scale and high verification difficulty. In order to ensure the safe operation and economic benefits of nuclear power plants, RPS needs not only reliability and stability, but also high availability. In order to explore the feasibility of verification of DCS availability index, based on Hualong One safety DCS engineering prototype, a set of availability testing methods and basic guarantee conditions suitable for DCS suppliers are summarized for discussion and exchange.

MTS (maintenance and test software), as an important human-computer interactive tool in DCS of nuclear power plant, plays a crucial role in configuration and management of the engineering applied data, engineering design, system debugging and operational maintenance. With the advent of information age, industrial designers pay more and more attention to interactive design, and user attaches increasing importance to the human-machine interactive experience of products, which puts forward new requirements for the human-machine interactive interface of MTS. This paper takes the human-computer interaction visual design of MTS as the core, through analyzing the human-computer interaction model and requirements of MTS, formulates the overall design scheme of the human-computer interaction system of MTS. According to the overall scheme, a set of human-computer interaction visual design method applied to MTS of DCS in nuclear power plant is proposed. The method and criterion of human-computer interaction design proposed in this paper is very important for the design human-computer interaction of nuclear power plant application software, and also points out the direction of human-computer interaction optimization of MTS mentioned in this paper.

In nuclear power plant embedded software control area, graphical control algorithms are described by Lustre program and designed with graphical editors of engineering station. After they are translated into C programs through a code generator, then the C programs are downloaded to the embedded platform software of NASPIC. The correctness of the code generator will directly determine the credibility and safety of NASPIC. Therefore, it’s important to ensure the correctness of the code generator translation process. Formally verified the correctness of a code generator, which has been studied for many years and becomes a most effective approach to develop a trusted code generator. This article is based on the requirements of nuclear safety-level I&C (Instrumentation & Control) system and mainly presents the development of a formally verified code generator front-end. The front-end translates Lustre source program into S_Lustre (a sequential intermediate language). And it is programmed and proved in Coq (Auxiliary Theorem Prover). During the development of the front-end, two key technical challenges have been resolved. The first is to ensure every id which appeared in the Lustre program has been defined. The second is to implement the casualty analysis and determine the evaluation order of user-defined types and equations based on the dependencies.

DCS (Digital Control System) is a very important part in safety protection in NPP (Nuclear Power Plant). Virtual DCS realizes the algorithm configuration of physical DCS running on general computers and gets rid of the limitation of hardware. In this paper, an emulation system (virtual DCS) is designed using virtual DPU technology. The virtual DCS can be used to verify the algorithm logic with the closed-loop system, which is composed by virtual DCS and digital reactor processing model together. The virtual DCS and controlled object are closed to test the control effects and in this way to verify the design of the NPP DCS. For the purpose of scientific research, training and design verification, virtual DCS should have malfunction simulation and other functions. These functions reflect the superiority of emulation technology.

The algorithm for the 1E system of NPPs is generally developed by a graph making tool at the engineering station. After the executable file is generated by the compiler and downloaded to the platform software, the credibility of the executable algorithm depends on the credibility of the generated programming code. Therefore, its security requirements put high demands on the compiler. The formal method has been developed rapidly in response to the problem of trusted compilation in recent years. This paper studies the main methodology and key technological components for designing and implementing the trusted compilers, meanwhile emphasize its industrial application based on existing research. The application is optimized, and a new composite compilation architecture is proposed, which reduces the difficulty of proof work and enhances the flexibility of the compiler’s later phase. Finally, several compiler optimization procedures for the security level software NASPIC are studied, and the algorithm compilation is enhanced, as well as the trusted performance of the device.

The translation-based virtual DCS based on Simulink can realize the function of virtualizing the physical DCS. This paper presents a method to virtualize physical DCS by using Simulink to realize the functions of running, freezing, acceleration and deceleration, setting and acquiring variable values, malfunction simulation, snapping and loading initial conditions, etc. Virtual DCS realizes the configuration translation by translating the configuration files of the physical DCS platform to Simulink platform, generating codes by using the configuration, and compiling the codes into algorithm files. The virtual DCS is combined with process model software to form a closed-loop, which simulates the level control system of the steam generator of a megawatt pressurized water reactor nuclear power plant for verification. Effective results have been achieved, which proves that the translation virtual DCS can meet the simulation effect of the simulator.

The pressurizer is an important equipment to maintain the stability of primary circuit in nuclear power plant. Pressurizer control system needs to be verified by effective testing. The most popular testing verification method on digital control system (DCS) is the discrete script tests. However it cannot make effective tests on the continuous variable process system. In this paper, a completed testing structure aiming on DCS is designed. The structure includes four levels: discrete script test, simulation test, stimulation test and online comparison test, which are in the progressive relationship, forming the comprehensive design and test system for the DCS. The system is applied to the verification of pressurizer in nuclear power plant. The HPR 1000 MW PWR model and pressurizer control system are used to make the simulation and verification tests.

The technology of Internet of things based on RFID has begun to play an increasingly important role in modern communication. according to the characteristics of system communication is an emerging networking technology and its application in modern communication system, the research analyzes the intelligent power grid planning and construction, according to the needs of modern communication technology, play advantage, put forward the application scheme of the Internet of things in power communication system, construction of basic network platform can carry out various types of services using EPON optical communications, wireless communications and other modern communications technology (network layer), carrying a variety of business data, voice, video and so on; combined with the technology of RFID, realize the communication management of RFID substation integrated monitoring system, the spare parts warehouse management system, computer room management system based on RFID. RFID computer room intelligent management system using RFID automatic identification technology and network and software platform of integrated equipment management system and the background of the front end of the RFID hardware equipment, equipment data room will focus to the control center is set through the network transmission, and storage, processing, real-time monitoring data to show, cabinets and other important business assets monitoring to provide faster, more carefully, more accurate daily management. From the point of view of the perception layer and network layer of the optical communication technology (optical fiber communication and wireless communication) the current status and Prospect of application in the Internet of things.

With the application and development of DCS control system technology in nuclear power projects in recent years, Fieldbus technology is gaining more and more attention of power generation groups. However, the application of fieldbus technology in the field of nuclear power is only at the primary stage. The main reason is that the requirement of high reliability of Nuclear Power Plant limits the development. This paper analyses the structure and typical faults of the fieldbus control system based on PROFIBUS technology, identifies the potential issues of the PROFIBUS fieldbus control system in conventional island of nuclear power plant and proposes a method to enhance the reliability of the fieldbus control system. Specific requirements are proposed on the redundancy design, diversity design, bus network segment design, bus equipment procurement, bus equipment installation and other aspects so that possible solutions could be provided for the reliability design and application of fieldbus systems in nuclear power plants.

With the business expansion, Nuclear Power Design Company has established a design portal based on the whole product line and integrated design platforms and tools. This indicates that nuclear power is changing from “document-centered design and production mode “to” data-centered digital design and production mode”. Meanwhile, the mass production of nuclear power has created an urgent application demand for a new form of archive management, information mining and design knowledge reuse in the digital environment. By analyzing the spatial evolution of archive object space in the nuclear power data state environment and using the concept of “three states” of archives in academic circles, this paper proposes the management idea of establishing the nuclear power archive object space, and reconstructs the nuclear power design process (knowledge) reuse model with “design activity” as the core and WBS (work break structure) coding structure as the framework.

The Sodium-cooled Fast Reactor (SFR) has become the research hotspot and been under development rapidly because of its small size and high power density in recent years. SFR contains a total of three circuits, primary sodium circuit, second sodium circuit, and third water circuit. The third circuit is feedwater supply system to cool the temperature of sodium in the second circuit during once-through steam generator (OTSG). Therefore, OTSG is considered as one of the most important equipment, which is essential to normal operation and safety assurance. In this article, the dynamic model of OTSG is built by the movable and fixed boundary method based on the conservation equations. Besides, the other components are also modeled to build the third water circuit model, such as the paralleling channel model, feedwater pumps model, and high pressure feedwater heaters model etc. In the MATLAB/Simulink environment, the third water circuit simulation platform is developed. The steady state condition and some typical disturbance conditions are simulated to verify the provided model. According to the dynamic characteristics and control requirements, the feedforward & feedback control system is designed to adjust the flowrate in time to insure the outlet temperature of sodium in secondary circuit within the prescribe limits. To verify the effectiveness of the provided control system, some typical conditions are simulated. The simulation results showed that the developed model can simulate the steady state and transient characteristics correctly, and the provided control strategy can achieve the control objective of a stable outlet temperature of sodium in secondary circuit. The research could provide the reference for the study of the control and operation for the OTSG and feedwater supply system in SFR.

The γ dose-rate monitor system of nuclear power plant is mainly composed of ionization chamber, pure aluminum electrode, armored cable, electrical penetration, and local processing box. During the LOCA environmental test of HPR1000, the output current of γ dose-rate monitor system ionization chamber detector decreases gradually and fails to return to the normal working range after being stable, which does not meet the requirements of the technical specifications. It is found that after the gas filling and sealing welding of the ionization chamber detector, the protective cover can't be installed due to the long inflation nozzle. The commissioning personnel does not take protective measures for the root when bending the inflation nozzle. Detection method is not used to verify the leakage point effectively, which results in the leakage point at the root of the inflation nozzle. Some suggestions are put forward, such as improving the processing assembly process and welding methods, ameliorating the testing methods and equipments, and enhancing the nuclear safety awareness of the commissioning personnel.

Software V&V technology provides reliability and safety guarantee for digital system and intelligent equipment which used in nuclear power plant. Although a systematic scheme of software V&V technologies has been formed for nuclear power plant, there are some technical nodes need to be discussed and considered. The software V&V standard has been updated. The higher requirements are put forward for the depth, scope, the appropriate techniques and methods of the software V&V. This paper will discuss the upgraded version of software V&V related standards, and give the key points that software V&V technology needs to pay attention. Then the pre-development software identification, new software V&V technical of the special instrument and control system, and the applicability assessment of the safety analysis software are also been discussed. The technical difficulties faced by software V&V in the above aspects will be put forward. At last the future development direction of software V&V technology is proposed.

The key technologies such as accident monitoring, installation and positioning, and signal abnormality of Nuclear Instrumentation System (hereinafter referred to as RPN) is studied in this paper. Through analyzing the requirements of post-accident monitoring laws and regulations, the compensation ionization chamber monitoring technology and fission chamber monitoring technology are deeply studied. The positioning mechanism of detector components is analyzed, and the positioning technical characteristics of “push-pull trolley type” and “bucket type” are systematically studied. The signal abnormality phenomenon is analyzed, and the equivalent circuit diagram and fault tree are established to analyze the technical causes of signal abnormality. Through the research on the key technologies, an advanced system scheme is independently realized. The independent scheme adopts fission chamber accident monitoring technology, realizes “bucket type” installation of detector components, and adopts measures such as improving gateway redundancy to ensure reliable and safe operation of the unit. The autonomous scheme has been applied to an advanced third-generation pressurized water reactor in China, and the research results have certain reference value for the research and design improvement of other reactors.

Rod Position Indication System (RPI) is the core component of Reactor Control-Rod Control System (RGL) in nuclear power plant. During the commissioning, periodic test and preventive maintenance of nuclear power plant, the performance test of rod position detector must be carried out. The resistance and insulation test results are important data to reflect the state of rod position detector coil and its related cables. Traditional test equipment needs external instruments and manual operation, which has the disadvantages of single function, low efficiency, poor data consistency and easy to produce human error. According to the heavy workload and short duration of RPI performance tests in the Nuclear Power Plant (NPP) commissioning, periodic tests and preventive maintenance, this paper designs an intelligent test system for RPI performance test in Nuclear Power Plant based on RPI test functional requirements analysis. The design meets the requirements of RPI resistance and insulation performance tests. In addition, the design is provided with an efficient, reliable and intelligent advantage and adapts to the complex working environment in Nuclear Power Plant. This design greatly facilitates RPI performance tests in NPPs.

The reliability of I&C system and software is critical to the nuclear power plants. As the most effective method to ensure the reliability of I&C systems, the quality, effectiveness and efficiency of V&V activities is a key issue. In order to reduce the human factors and improve the efficiency and quality of verification, an automated documentation structural integrity method is proposed based on natural language processing. Sample tests are performed to prove the effectiveness of the proposed method.

As an important process control system in nuclear power plants, non-safety digital I&C system has the characteristics of numerous internal and external interface, complex functions and low development integrity level. Users pay more and more attention to the quality in each phase of its life cycle, such as design, equipment manufacturing, commissioning and verification, delivery, operation and maintenance. This paper studies the software quality evaluation process, and proposes a set of applicable regulations and standards system and quality evaluation model. The contents and key elements of quality evaluation in three important phases of life cycle process are discussed in detail, and the V&V technical requirements and quality management requirements are integrated, which provides a reference for the implementation of new building nuclear power plants and modernization projectsPer.

Digital instrumentation and control (I&C) system has been widely applied on large commercial pressurized water reactors (PWR), which brings significant convenience to their operation. However, the level of automation and operation efficiency of nuclear power plant is still lower than conventional power plant such as thermal power plant or hydro power plant. On one hand, the safety and reliability requirement of nuclear power plant always chooses the mature and proved technology, but not the emerging technology such as artificial intelligence. On the other hand, the complexity of system design and operation requirement also makes it difficult to improve the automation level of nuclear power plant. To deal with these difficulties, a statistical analysis has been given to identify the challenges on improving the automation level. And a framework of plant level autonomous is given to illustrate how to organize the techniques and algorithms available to reach this goal. For the new functions, algorithms and equipments that involved in the framework, the requirement of nuclear safety should be considered. Therefore related requirements that need to be addressed are also discussed in this paper.

As a diversified means of microprocessor technology, FPGA technology has been adopted in an I&C system important to safety independently developed by a Chinese nuclear power group due to its advantages of reducing the overall complexity of NPP I&C system and effectively solving the hidden dangers of DCS common cause failure. Internationally, FPGA V&V technology for NPP I&C system important to safety, is still under discussion and has not yet formed a consensus operating standard. However, the domestic FPGA regulations and standards in the field of nuclear power are not perfect, and the engineering practice experience is insufficient. All kinds of V&V technologies, methods and their engineering applications are being explored and improved. Therefore, it is urgent to solve the engineering application problem of FPGA V&V technology in digital systems that perform safety important functions in the field of nuclear power. Through the research on relevant regulations and standards of FPGA V&V for NPP I&C system important to safety, A FPGA V&V general technical scheme and method system suitable for NPP I&C system important to safety is established, and the application effects of the scheme and method is tested in project practice. Finally, the technical key points in the process of FPGA V&V for NPP I&C system important to safety are summarized.

As the nerve center of nuclear power plant (NPP), distributed control system (DCS) has been widely used with perfect control function, flexible system configuration, safe and reliable operation state and strong applicability. The relevant modification or upgrade of DCS is the focus of NPP's digital modification or upgrade. The reliability of the computerized system or equipment of DCS depends to a large extent on the reliability of the software, and systematic failure may be introduced in the development process of the software. Software verification and validation (V&V) technology is recognized as one of the key technologies in the field of nuclear power that can effectively ensure and improve the quality of software. Software V&V technology for safety DCS system has been widely studied and applied, but software V&V technology for non-safety DCS system has not attracted attention from all parties. Based on this, through studying NPP's non-safety DCS software V&V related standards and technical reports, combining with the analysis of NPP's operating experience in operating event reports, a V&V scheme based on system engineering method is proposed, and the activities and tasks of each phase of the scheme V&V process are preliminarily summarized, and the detailed guiding standards or technical reports that can be referred to when executing V&V tasks are expounded. At last, the research results are expected to provide technical reference for the modification or upgrade of non-safety DCS in software.

Safety digital control system (DCS) cabinet, as a carrier for the electronic devices, plays a significant role in ensuring the normal operation of the nuclear power plant. In this work, a design idea of impact resistant cabinet frame with high strength and low quality has been proposed. It adopts the design ideas such as flat frame section design, distributed multi-column layout, integrated structural and functional design and preservation of a certain safety margin, which can be used to guide the design of cabinet with high stiffness and strength, small column size and low weight. In addition, finite element analysis and impact response analysis of the designed cabinet are carried out by using ABAQUS-DDAM method. Results show that the designed cabinet is characterized by high stiffness, high strength and low weight, and the cabinet possesses good ability in resisting high level of impact. The design ideas proposed in this research can provide reference for the design of structural parts of other anti-shock electronic equipment.

This paper introduces a new method of research and application of software V&V based on embedded technology in Nuclear Power Plants. Based on software abstract test, the relationship between defect trend and cost is analyzed. Mind map test analysis is carried out, and the V&V levels are carefully defined and divided. System tests based on model are carried out. Specification-based tests and code-based unit-level tests are organically combined. Finally, this paper summarizes the V&V scheme. This V&V method based on embedded technology has been unanimously approved by experts and has a widely application prospect.

This paper studies the quantitative method of Time-PSF in HRA of nuclear power plant, and determines the precondition of quantitative calculation, and combs the quantification process, also introduces the key steps of quantification process in detail. Taking the “Medium Break LOCA (superimposed cooling failure)” event as an example, the processing flow based on SOP (state oriented operation procedures) and SEOP (symptom oriented operation procedures) emergency response procedures is analyzed respectively, and the time required is quantified by the method introduced in this paper. The method proposed in this paper solidifies the process, and ensures that the analysis has reasonable preconditions, and solves the problem of difficulty in obtaining basic data and difficult to quantify, this method makes the Time-PSF assessment more accurate and reliable on the basis of quantitative calculation, and then improves the accuracy of evaluating the probability of human error (HEP) in the digital main control room, so as to provide reference for HRA staff. In addition, through the implementation of the calculation method, it is helpful to obtain relevant actions with high probability of human error, and further optimize the function allocation and system design.

The press-in terminal is a special-shaped cross-section suitable for press-in connection. According to the different design of the press-in connection, the press-in connection is divided into two processes: solid press-in connection and flexible press-in connection. This paper mainly analyzes the requirements of the PCBA flexible solderless press-in connection process introduced in the DCS safety equipment of nuclear power plants, identifies the key control requirements. The test items are determined by the design of the application scheme, and various forms of tests are conducted. The data of test is obtained. Based on the test results analyzed, the reliability of the flexible press-in connection process implementation is verified, which laid the foundation for the application of the later flexible press-in connection process, and can also form a set of evaluation methods for the flexible press-in connection process.

In the past few years, human-caused failures have occurred frequently at the nuclear power plant, such as the wrong forcing of signals, the wrong interval, and improper operation. They all have directly produced serious consequences. In the physical protection system, security personnel misoperations occurred in multiple projects, resulting in varying degrees of damage to the equipment. Therefore, how to effectively improve the reliability and effectiveness of the civil air defense of the physical protection system becomes particularly important. This article uses Human Factors Engineering to realize the optimization of the physical protection system in terms of environmental design, work area layout, equipment maintenance space and alarm management, so that security personnel can improve work efficiency in a comfortable environment, and achieve the best between man, machine and environment.

The Long-Short Term Memory (LSTM) model is applied to the Steam Generator (SG) water level prediction in this work. The model is designed and implemented within the SIMULINK environment, where a real-time validation platform is also constructed using traditional PID controller. The SG water level feedback signal is fed to both the LSTM model and the PID controller, allowing the prediction to be generated online and compared to the actual controlled value. The results have demonstrated the functionality and advantages of the LSTM model, such as the high accuracy with prediction error of −1.1887 × 10–4, low loss value with MSE of 1.4130 × 10–8, and quick convergence during the simulation. Discoveries found in this work could enable future exploration of developing deep-learning-based control strategy for nuclear power plants.

The wide application of information and network technology in the field of nuclear power not only brings positive influence to the development of nuclear power, but also brings new threats and challenges to Nuclear Safety. In order to reduce the potential risks brought by network attacks to nuclear power units and guide Nuclear Power Plant to carry out information security design, this paper studies the defense-in-depth requirements related to Instrumentation and Control system information security standards. This paper firstly compares domestic and foreign laws and standards combined with the current situation of China's Nuclear Power Plant information security laws and standards, and then the paper analyzes the relevant contents of different laws and standards involving defense in depth. Finally the paper provides reference for the construction of China’s nuclear power information security system.

Before the implementation of the extended fuel cycle improvement in nuclear power plants, it is necessary to conduct an adaptive analysis of the relevant systems and equipment cycles. The purpose is to show that the improved cycle still meets the requirements for safe operation of nuclear power plants. This paper firstly introduce some problems: How to demonstrate the equipment cycle affected by the refueling cycle, and what methods can be used to analysis many type of I&C equipment’s cycle which are manufactured by suppliers in nuclear power plant 18-month refueling. Based on the engineering experience of extending the refuelling cycle of domestic and international nuclear power plants, this paper briefly introduces several periodic demonstration methods and explains their characteristics at the same time, which provides guideline and practices for the nuclear power plants that implement the fuel cycle extension plan.

In recent years, the potential spurious actuation posed by digital I&C systems has obtained considerable attentions in nuclear industry and the associated methodology development has become an important and urgent issue. A new identification method for spurious actuation of digital I&C systems is developed. It is a top-down process which begins with a basic identification of the potential spurious actuation event and then associated influence is evaluated if the spurious actuation could challenge plant safety. Subsequently, the identified spurious actuation of I&C functions are analyzed following the safety analysis. As an example, a special spurious actuation event is defined and analyzed.

At present, digital control technology is widely used in safety I&C system of nuclear power plant. In addition to improving the self-diagnosis coverage of the equipment to increase the reliability of the system, periodic tests are also widely used, but how to determine the test interval of the safety I&C system in theory has gradually become a new research direction. Base on this, this article models the periodic test interval from the beginning of design and combines with the PSA technology by analyzing the initial events of a typical NPP, and integrates it into plant PSA models. Through the analysis of calculation results, the weakness of the test interval is identified and the improvements are given. After that a set of design process for periodic test interval are summarized and ultimately enhance the overall safety and reliability objectives of NPP.

Nuclear instrument system (RPN) continuously monitors the neutron flux rate level and the core power level in the reactor core and participates in the operation of the reactor protection system, which is a key instrument and related to reactor safety. In the past two years, Nuclear Instrumentation (RPN) Source Range signal abnormal fluctuation events have occurred occasionally in multi-base projects. Studies have found that electromagnetic interference (EMI) is a main cause of signal abnormal. This paper discusses adding electromagnetic compatibility design from the top-level design. Through equipment design to meet the requirements of RPN signal detection equipment and cables on electromagnetic environment, improving the stability of the signal link by installing technical requirements, through equipment identification and other means to verify the cabinets' requirements for electromagnetic environment, summarize the good practical experience of the power plants in operation, formulate systematic and detailed EMI protection requirements and implementation measures for RPN system, to avoid signal flash and improve the stability of the power station.

In the early 1990s, PLC replaced the conventional relay control circuit, which was popular in various fields, and greatly simplified the control circuit of power plant. With the development of technology, all kinds of automation technology has been widely used in various industries. DCS gradually replaces PLC, which inevitably leads to various problems in practical application. This paper mainly from the controller configuration, logic function design and redundant fault-tolerant mechanism, introduces the problems in the transformation of the status quo in details. Through the comparative analysis and qualitative research methods, the paper analyzes the design of important functional interfaces of turbine trip and rejection in Nuclear Power Station, and gives solution to the problems of transformation.

In the process of construction and commissioning of million kilowatt pressurized water reactor (CPR1000), due to design optimization or technical parameter adjustment, there will be a certain amount of technical changes. Technical changes are generally completed and closed by different departments of the construction unit through four steps such as design, distribution, implementation and reappraisal. The process of technical changes is complex. The amount of coordination work between relative departments is large, so that the technical changes have shortages of reissue, missing, missing implementation, missing reappraisal, etc., which grows the risk of the safe and stable operation of the unit. By studying the optimization of the management systems, the management systems would found the functions of automatic screening and filtering, which can automatically form the implementation and identification tasks. Furthermore the systems found relationships between the upstream and downstream, realize the visual query and tracking of the upstream and downstream related process, and realize the closed-loop management of the technical change process.

In order to ensure the safe and stable operation of nuclear power plant (NPP), the safety reliability analysis is necessary. The safe spray control system is an important engineered safety features system which is use in the accident about containment over-temperature, over-pressure. The article gives an research on the system theoretic process analysis (STPA) methods for the reliability analysis of the safety spray control system function. It helps to find out the potential factors effect for the safety function of the spray control and provides some suggestion to improve the system reliability and the station safety operation. This is the good experience and new demonstration for the STPA application and is benefit to the following project practice in NPP.

The reliability of the operator's response process after a nuclear power plant accident has an important impact on the overall reliability of accident mitigation. The automatic diagnostic function of nuclear power plant state is designed for the advanced digital control system, by monitoring and processing the plant, it provides an initial orientation or reorientation diagnosis for the Emergency Operating Procedures (EOPs) during emergency operating conditions. Therefore, the operator response process in main control room, especially the potential human errors have some new characteristics, when compared with the traditional way. The qualitative assessment of the operator response process forms the basis for the quantification of the associated Human Error Probability (HEP). The purpose of this paper is to study the reliability of the operator actions required to establish simultaneous Hot Leg injection following an Intermediate Break Loss of Coolant Accident (IBLOCA), a typical accident condition of nuclear power plant. The accident sequence and operator's actions are given, the detail qualitative and quantitative assessment are implemented base on the foundation of SPAR-H method by a constructed fault tree. The result shows that the failure probability of operating from the Auxiliary Control Panel (ACP) is higher than operating from the Plant Computer Information & Control System (PCICS). The main recommendations are providing more training for operation from the ACP following a loss of PCICS, increase descriptive information within the EOPs and the Human Machine Interface (HMI), providing a dedicated plant status display system and then decrease the reliance placed on the knowledge and memory of the operators to understand important information about plant configuration. The reliability assessment helpful to improve the human factor suitability, provide guidance for optimize the operator's response process and effectively improve the reliability of engineering design under an IBLOCA accident scenario.

Turbine speed-up control is the key part of power unit start-up. In this paper, the sequential control of one-key start of a nuclear power steam turbine is studied. It is found that the false grid-connected signal is one of the potential risk factors that lead to the failure of turbine speed-up control. An optimization scheme of increasing false grid-connected signal detection and automatic locking speed-up is proposed, which improves the reliability and safety of unit speed-up control and provides a reference for subsequent unit one-key start-up design.

Humidity regulation and control of ventilation system in nuclear power plant plays an increasingly important role in the third generation nuclear power plant. Humidity meter is an important equipment to realize this function. In this paper, the measuring principle of humidity instrument and the application requirements of the third generation nuclear power humidity instrument are compared and analyzed in detail. Combined with the application of humidity instruments in nuclear power plants, two kinds of humidity instruments, capacitive and optical fiber, are proposed to meet the application requirements of the third generation of nuclear power plants.

Traditional paper-based procedure play an indispensable and important role in guiding the operation and management of nuclear power plants. However, with the development and application of full digital instrument control technology and advanced main control room, its shortcomings such as low execution efficiency and mismatch with advanced technologies are beginning to emerge. In order to take full advantage of computers, most nuclear power country has developed the computer-based procedures (CBP) system itself. This paper analyzes the requirements of regulations and standards in domestic and foreign country, compares and explains the technical points of typical nuclear power units in terms of functional level, computerized range, procedure interface, automation level, upgrade maintenance, etc., summarizes the advantages and disadvantages of current CBP system, and puts forward the optimization and improvement technical scheme for future CBP application.

As one of the most sensitive elements of human vision, color is widely used in Human-System Interface system. The quality of color system often plays a vital role in the success or failure of interface design. Its effect directly affects the user’s operation experience. Therefore, nuclear power plant Human-System Interface design needs to establish a complete color system. Based on the color system of CPR1000 series power plants and the operation experience feedback of in-service power plant operators, the Human-System Interface color system of HPR1000 reactor type is proposed, and the compliance analysis of the regulations and standards is carried out, in which the relevant requirements of international and domestic laws and standards for color use are analyzed in detail, and the color use schemes of various reactor types in-service power plants are compared.

Study for the reliability assessment for the important human actions which are required after reactor core damage in Severe Accident (SA) caused by Large Break Loss of Coolant Accident (LB-LOCA) is carried out in this paper. These human actions are analysed by qualitative assessment and quantitative assessment. The reliability of reactor pit injection measure in this scenario is very important so the Performance Shaping Factors (PSFs) which may cause the failure of the mitigation measures are identified and assessed. Based on the assessment result, three improvement recommendations are made to improve the success probability of protecting Reactor Pressure Vessel (RPV) integrity in SA.

The Plant Computer Information and Control System (PCICS) is the main digital means for operators to monitor and control the Nuclear Power Plant (NPP), providing operators with centralized operation control, information display, comprehensive processing and data storage functions. The PCICS plays an important role in the operation of NPP. The objective of PCICS life monitoring is to make sure that in case of failure of PCICS, a hardware alarm can be triggered so that the operators will be informed and perform the appropriate actions. Basing on the equipment classification and function analysis, this paper analyses the equipment and functions that should be included in PCICS life monitoring scope, formulates effective countermeasures. The paper finally completes the implementation scheme of PCICS life monitoring with many innovations and advances at the same time. It is very meaningful for the PCICS system because it comes up with clear logic and safe structure for the availability surveillance of the main control and monitoring mean of NPP.

Large rotating machinery is very extensive and important equipment in nuclear power plant. It is necessary to carry out on-line vibration monitoring and effective fault diagnosis, which are also the basic application characteristics of intelligent nuclear power plant system. Support vector machine (SVM) can be applied to the fault diagnosis of large rotating machinery in nuclear power plant because it can achieve better classification effect with less training samples and no prior knowledge of fault classification. Therefore, firstly, an online vibration monitoring system for large rotating machinery in nuclear power plant is constructed to extract fault features. Then, according to the simulation data of vibration test-bed and the fault data of actual equipment operation, SVM algorithm is used to simulate. By comparing with the actual fault, the diagnosis result is better, which verifies the effectiveness of the method, and is intelligent for the software platform of vibration monitoring device Diagnostic function development provides support.

This paper analyzes the main failure modes of fire protection system in nuclear power plant, and uses the FMECA (failure mode effect and criticality analysis) method to finds out its key sensitive equipment. Combined with the working principle and operation environment of fire protection equipment, it analyzes the failure modes of key equipment in depth, and proposes the targeted operation and maintenance optimization strategy, so as to reduce the failure probability of fire protection system equipment and improve the safety and reliability of fire protection system in nuclear power plant.

As an essential class 1E primary instrument of the core cooling monitoring system (CCMS), in-core coolant level detector is used to monitoring the coolant level inside the reactor pressure vessel (RPV). Normally functional test is fundamental means to verify whether the in-core coolant level detector can fulfill its safety function. However, due to the inappropriate testing approaches and the incomplete design, the level tracking curves of the in-core coolant level detector developed for HPR1000 are unsatisfactory during the functional test, which shows a certain degree of deviation from the expected curves and could not track the real coolant level variation in time. After technical analyzing, the root cause of the issue is identified, and the improvement in test methods and the level detector’s structural design is implemented. Then the improved level measuring curves of the in-core coolant level detector are shown to prove the effectiveness of the measures having been taken.

This article mainly discusses the application of intelligent IP camera in the video monitoring system of nuclear power plant. Firstly, three monitoring requirements of existing video monitoring system of nuclear power plant are analyzed, and the advantages and characteristics of intelligent IP camera are given. On this basis, the monitoring scenarios that can be realized by the video monitoring system of nuclear power plant based on IP camera are given. It is concluded that the application of intelligent IP camera can greatly improve the digital and intelligent level of nuclear power plant monitoring system.

With the development of artificial intelligence technology, recently the concept about smart power plant has been discussed widely, and also has been used in a lot of industries even nuclear power plant. Chinese heavy-duty gas turbine (CHGT) development and application (D&A) is one of national science and technology major projects and being implemented by China United Gas Turbine Technology Co., ltd (UGTC). According to the progress of CHGT D&A, this paper firstly points out the framework how to build smart power plant and digital twin in the process of CHGT D&A for different phases, such as engineering phase, manufacture phase, construction phase, commissioning phase, operation and maintenance phase. Secondly the paper describes two building modes which are forward and backward modes, and then the contents have been discussed based on the different conditions in practice of the power plant. Finally taking backward mode as an example, the paper provides a feasible technical solution. Therefore it would be benefit for the future research about smart power plant for CHGT D&A.

Based on the convolution neural network (CNN) technology, and the use of video monitoring equipment, an intelligent safety monitoring system for nuclear power plant is designed and developed, to realize the continuous monitoring of regional fire situation and whether the personnel wearing safety helmet or not. This paper introduces the framework of the system, the principle of convolution neural network and the feature recognition strategy, and describes the process of system training data acquisition and model training in detail. The 6-month continuous measurement in nuclear power plant shows that this system is stable, accurate and timely, and can effectively improve the efficiency of field monitoring.

Based on the analysis of the high temperature high pressure heater level switch fault in the Conventional Island of the 1000 MW nuclear power unit, this paper adopts optimization of instrument selection by replacing level switch with level transmitter. By comparing the working characteristics of three most commonly used transmitters, guided wave radar liquid level transmitter is finally selected to improve the measurement reliability and stability of unit operation. The application of guided wave radar has provided the reference engineering experience for subsequent new projects or in service transformation.

The number of operating units of nuclear power plants in china has gradually increased, but foreign nuclear equipment suppliers have gradually decreased. Commodity grade item assessment has played a very good role in the procurement of unique spare parts and strengthening the level of equipment quality control. In order to verify that the equipment can complete its predetermined functions with the specified technical performance under the operating conditions and environmental conditions specified on the site, the identification test of the equipment is an important part of the commodity grade dedication. In this paper, the qualification test of the level transmitter used in the 1E diesel engine lubricating oil system of the second-generation improved nuclear power plant is studied. This paper analyzed the versatility and differentiation requirements in IEEE 323, RCC-E, KTA3505 related equipment qualification standards, introduces the test content and results, and provides technical reference for the applicability of commodity grade dedication.

The load following requirement will be increased greatly while the Small Modular Reactor (SMR) is used for isolated island operation mode. In order to improve the load following ability of the system, the flexible load (FL) is presented in this study to follow the electricity load during load follow. The definition of FL is a type of power consuming system that can change the needed power discretionary. For example, the hydrogen manufacture plant, the desalination facility, the district heating system, and other industrial process heat supply system are all FLs. In order to analyze the dynamic characters of the system, pressurized water reactor (PWR) type SMR (comprised with reactor core and once-through steam generator), balance of plant (BOP), and FL are modeled first. Based on these dynamic models, a SMR-FL dynamic simulation program is built in MATLAB/Simulink environment. According to different system characteristics, two types of FL are considered, and the related load allocation control systems are designed. A typical electricity load is used to verify the designed control systems for the two types of FL. The simulation results show that the designed control systems can achieve load following by adjusting the power consuming of FL and maintaining a constant or moderate changing output of SMR. Obviously, this character can make the SMR-FL system more flexible and efficiency.

This paper takes verification of the reactor RRC system of a pressurized water reactor nuclear power plant as the research object. This paper determines the modelling scope of the RRC system according to the characteristics of the RRC system, modeled its process systems and control systems, and eventually constructed the digital Twins system of the RRC system, which realizes the dynamic closed-loop verification of the RRC system. The paper gives examples of implementation for typical accidents. The research results eventually achieve the purpose of accurate verification of the RRC system, and improve the safety and economy level of nuclear power plants, and play an important role in improving the safety and economy level of nuclear power plant operation.

The response time of shutdown of a nuclear reactor is an important parameter of reactor protection systems in nuclear power plants. Furthermore, as a part of a reactor protection system, the point to point communications system mostly influence the response time of shutdown of a nuclear reactor. A new solution to optimizing the response time is introduced, where a point to point communication system is implemented by FPGA technologies. The new designed system could extremely improve the efficiency of communications, which contributes to shortening the response time of shutdown. This paper delivers not only analysis of the response time of shutdown in current systems, but also the design scheme of the FPGA-based system. In addition, the communication system based on FPGA optimizing the response time of shutdown of a nuclear reactor is proved which is applied in reactor protect systems of VVER.

With the development of digital and intelligent technology in industrial control, the safety and reliability of control and protection system is becoming increasingly prominent. As the Pearl of modern industry, heavy duty gas turbine is very important. The ongoing independent R&D work of UGTC (China United Gas Turbine Technology Co., Ltd.) has been highly concerned by the industry. As the nerve center system of gas turbine, whether the control and protection system can complete the expected function safely and reliably is also an important part of the independent R&D work of heavy-duty gas turbine. In this paper, combined with the independent research and development experience of nuclear power control and protection system in China, based on the functional safety certification requirements of IEC 61508 standard, the technical scheme of heavy-duty gas turbine control and protection system software V&V is discussed, and the relevant technical key points are put forward, which has a certain guiding significance for the design and commissioning of control and protection system of heavy-duty gas turbine.