An essential material testing facility (EMTF) used for pebble bed effective conductivity measuring experiment of high temperature gas-cooled reactor was developed. As required, temperature measuring points were arranged in EMTF to monitor the temperature field. The maximum test temperature in EMTF would be 1600°C. The main part of EMTF was made of graphite and carbon felt, hence the measuring points were in strong carbon reducing environment, of which some were embedded in the carbon felt. The material choice and structure design of the thermal couples were introduced. In the high temperature region, our specially designed tungsten-rhenium thermocouple was used. For the comparison, some imported tungsten-rhenium thermocouples were also used. The distributive data acquisition system was made of IMP form Solartron in England, which with industrial computer constructs a data acquisition and control system. The software was programmed with configuration software of Sunwayland Forcecontrol. Finally, the measuring results in each part of EMTF were given, such as the heater, testing zone, thermal insulating layers. The experimental result showed that the designed temperature measuring system worked stably and reliably, which could finish the temperature acquisition task in 1600°C high temperature carbon reducing condition. After the experiment, the thermal couples experiencing high temperature in carbon reducing condition were removed out to analyze the structural change and study the damaged condition.

In this work, in-depth considerations about FPGA-CPU architectures design methods are presented together with some possible plant applications. FPGAs are used because of their potentials in guaranteeing parallelism, synchronism and high reliability tasks; on the other hand CPUs are crucial to higher precision computations and to reproduce data and results on a graphical interface, thus resulting in an improvement in the human-machine-interface (HMI). Design choices are here reported to justify the benefits from the use of both CPU-FPGA approaches. Taking advantage of these features, the LabVIEW environment and its digital platform is expanded in order to develop a novel tool able to support designers during the fundamental phases of a supervision and control tool realization: from the model development to the digital hardware implementation and testing. LabVIEW is able to communicate with several languages like Matlab or Scilab for a theoretical model study and simulation, and VHDL for a digital implementation, taking the advantages from both development environments. Main results have been reported concerning studies of control systems in nuclear plant facilities, showing how a full-digital platform can be very useful in developing monitoring and control instrumentation. This new tool is thought to improve the so-called hardware in the loop (HIL) simulations and to accelerate the process of prototypes implementation and testing, using a more accurate and reliable environment in terms of performance and safety.

Power-level regulation is a key technique that provides both economic and stable operation for any nuclear reactors. Since the pressurized water reactors (PWR) are not only the most widely utilized nuclear reactors but also complex nonlinear systems, efficient nonlinear power-level control strategy is meaningful to not only the development of nuclear power technique but also the renaissance of nuclear energy. Further, since self-stability is the most crucial dynamic feature of nuclear reactors, it is necessary to give self-stability analysis which gives the means to strengthen the closed-loop stability. In this paper, a sufficient condition of globally asymptotic self-stability for PWR is firstly given. Based on this self-stability analysis, a nonlinear state-feedback power-level control law with L 2 disturbance attenuation performance is then proposed. This new controller is applied to the power-level regulation of a nuclear heating reactor (NHR), and simulation results show its feasibility and high performance.

The product of Fangjiashan Nuclear Power Plant (NPP) Unit1 Fuel Handling and Storage (PMC) System is the first Chinese PMC system, with its own intellectual property rights. During the development and design lifecycle for the product, China Nuclear Control System Engineering Co., Ltd (CNCS) was responsible for Verification and Validation (V&V) processes of the PMC system, taking a significant role for ensuring the reliability and safety of the software in PMC system. The PMC V&V project is also the first ever independent V&V project of the China nuclear industry. V&V is a technical discipline of systems engineering. The purpose of V&V is to help the development team build quality into the system during the life cycle. The V&V processes determine whether the product developed satisfies its intended use and user’s needs. V&V provides an objective assessment of products and processes throughout the life cycle. This assessment demonstrates whether the requirements are correct, complete, accurate, consistent, and testable. CNCS V&V PMC project team performed PMC V&V activities in compliance with IEEE Std 1012-1998, which is endorsed by NRC Regulatory Guide 1.168-2004. After critical analyses by V&V team, the Software Integrity Level (SIL) of PMC is SIL3, so associated V&V activities must conform to the minimum tasks list in IEEE Std 1012-1998 Table 1. After 2 years of hard working, Fangjiashan NPP Unit1 PMC system V&V activities have been completed. The whole activities and methodology was passed the final evaluation by the customer and was highly compliment by the international V&V expert committee at year of 2012. This paper discusses how to establish V&V organization and infrastructure, and then how to perform V&V activities in high quality, and the specific new and best practice methods, approach and tools for executing V&V tasks. Because the PMC V&V project is the first ever independent V&V project of the China nuclear industry, the methods, approach and tools which was executed in the whole lifecycle and depicted in this paper could help to guide and be referred by the future V&V activities and projects.

LEFM (Leading Edge Flow Meter) Ultrasonic Flowmeters have been installed in feedwater pipes in over 68 Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear plants worldwide since the late 1970s. The 8-path, crossed-plane configuration commonly installed in feedwater pipes has been demonstrated through extensive calibrations at nationally certified calibration laboratories to provide feedwater mass flow accuracy better than ±0.3%, regardless of upstream hydraulic conditions. The increased accuracy has been accepted by nuclear regulators in several countries as the basis for an increase in licensed thermal power of nuclear power plants. LEFM technology provides the ability to methodically check all key contributors to flow measurement uncertainty, thereby providing continuous quality assurance of flow data provided by the system and assuring plant operators that the meters remain within their design basis. This paper will describe how LEFM chordal ultrasonic flowmeter operation and data quality can be verified, thereby providing the basis for troubleshooting changing balance of plant indicators. The paper will discuss a methodology for estimating changes in moisture content of the steam produced by steam generators at a PWR and will present a recent case study showing how detailed analysis of on-line feedwater flow data has been used to reconcile readings of other plant instrumentation to identify changes in steam moisture of ±0.1%.

Because of the inherent safety feature, the modular high temperature gas-cooled reactor (MHTGR) has been seen as one of the best candidates for the next generation of nuclear power plants. Power-regulation is a key technique to provide safe, stable and efficient plant operation for the multi-reactor-module power plants. Based on the successful design and operation of Chinese test high temperature gas-cooled reactor HTR-10, the high temperature gas cooled reactor pebble-bed module (HTR-PM) project is proposed. Since the HTR-PM is composed of two interconnected nuclear steam supply system (NSSS), the operation strategy of the multi-reactor-modular HTR-PM is different from that of those one-reactor-modular high temperature reactors such as HTR-10. In order to design and verify the operation strategy for HTR-PM, it is necessary to establish a real-time simulation platform. The design and development of this platform is presented in this paper.

The power system frequency stability problem has been the focus of attention since 1920s. Primary frequency regulation (PFR) is one of vital methods to maintain frequency stability, especially when the power system suffers a sudden load disturbance, such as the substation tripping or generator outage accident. To a power system, the more power units participating in PFR, the frequency stability will be controlled better. With the rapid development of nuclear power units in China, the study of the nuclear power units participating in PFR of power system has becoming a pressing problem. In the present study, a detailed nonlinear dynamic mathematical model of the whole pressurized water reactor (PWR) nuclear power plant is built. The dynamic calculation codes are compiled to dynamic link library (DLL) files, which are developed by using FORTRAN language, and the compiled DLL files are embedded into MATLAB/SIMULINK simulation platform by using S-function, the control systems are designed and the whole system of PWR nuclear power plant is simulated in MATLAB/SIMULINK finally. In this method, the operation and control mode of PWR participating in PFR of power system is analyzed and simulated. The simulations results show that the PWR nuclear units are feasible in participating in PFR from safety and economy.

The Enhanced CANDU 6® (EC6®) is a nuclear reactor based on the proven CANDU 6 reactor that uses natural uranium and heavy water as a coolant and as a moderator. The EC6 has many enhancements to improve production and to meet next generation reactor standards. A modern distributed control system is used coupled with a new plant display system to facilitate the operation of the EC6. Computerization of safety systems is extended to include neutronic trip parameters and the Emergency Core Cooling (ECC) system. Computerized systems allow for thorough verification during their development, minimizing the installation, commissioning and maintenance times. Testing of these safety systems is also computer assisted to reduce the time required for testing and reduce the possibility of spurious trips. Enhanced safety system monitoring and equipment monitoring systems are used for early detection of faults or deterioration of equipment. This can in turn allow more targeted maintenance activities and improved safety. Critical safety parameter monitoring is also included in the design. Fission chambers are used instead of ion chambers for measuring neutron power. Fission chambers can extend the bottom of the power measurement range by three decades, allowing outages and restart of the reactor to be performed without the installation of startup instrumentation. Instrumentation is included to cater for Beyond Design Basis Accidents (BDBA). Additional battery banks are also included in the design to cater for Station Black-Out (SBO) and extended SBO. Connection provisions for mobile power generators have also been included in the design. The safety systems are controlled by modern equipment qualified to international standards for safety systems. The instrumentation required for these systems is also assessed for the severe conditions to which they might be exposed. Lessons learned from Fukushima on protection from external events are included in the design.

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

High Temperature gas-cooled Reactor-Pebble bed Module (HTR-PM) Reactor Protection System (RPS) is a dedicated system to be designed and developed according to HTR-PM Nuclear Power Plant reactor protection specifications. HTR-PM RPS has the framework of four redundant channels and has two independent and diverse subsystem x and subsystem y to perform different protection functions, which would decrease the potential common cause failure caused by software and increase the system reliability.

With the application in HTR-10GT, the reliability and stability of the AMB system should be studied deeply. Especially EMI analysis on the switch power amplifier is needed to be done, since which one is the main interference source for AMB during the switch turn-on and turn-off. Based on it, a simplified and improved modeling method is listed by dividing the nonlinear transition into several stages, and the models of the voltage source and current source are built in the form of the piece-wise linear way. The conducted emission on the differential mode noise and the common mode is shown by simulation. The result could provide the theoretical basis on the designing the grounding, filter and isolation for the AMB system.

Due to development of electric power system, the proportion of nuclear power and its needs for Load follow operation have become large. Then there would be higher requirements to the power control ability of the nuclear power system. The project with China’s own brands and using the advanced pressurized water reactor (APWR) nuclear power technology, CPR1000, adopts a control method called Mode G, using G banks, N banks, R bank and soluble boron to perform the load follow control. The common Point Model, widely used in conventional core modeling, cannot successfully research the axial power distribution. But in this paper, a two-node dynamic function model was built, taking CPR1000 as an example, considering the coupling coefficient and mutual influence. The transient parameter values were obtained by the steady-state calculation of single channel with its original structure parameter and operation parameter. Then the core system simulation of CPR1000 was modeled. In order to carry out the load follow operation without the boron adjustment, known as BTP mode, the control banks were regrouped to control the average temperature by M banks and axial offset by AO bank. The last, the optimal control implementation was connected to the core simulation platform to operate with the variable daily load. The feasibility and effectiveness of the control policy and the simulation system are validated by the calculation results via one-dimensional in axial computer code APOC. Results of computations performed for a reference reactor were present, giving the possibility that the optimal control policy could stretches the capability of the reactor to follow an average daily load curve. The paper demonstrated in principle that the BTP mode was feasible for CPR1000.

In the nuclear power plant, the off gas linear monitor had been installed for the purpose of a detection of fuel leakage. However, its sensitivity to detect the early stage of fuel leakage was not enough, because it measured only the gross gamma-ray. So a high sensitive off gas monitor has been desired. HITACHI, Ltd., is commercialized the Off-Gas Nuclide Analysis System which can be accurately detected radioactive noble gases as a index to fuel leakage, and delivered it in 19 plants since 2002. It has been continuously working well and able to detect the early stage of fuel leakage 7 times. It can be detected the minute fuel leakage, so it is possible to operate the plant till next periodic check of the plant by executing PST*. And even if it is the case in which the plant must be stopped, the extent of pollution by fuel failure can be minimized. Main features of the Off-Gas Nuclide Analysis System are followings: (1) Maintenance-free continuous measurement is enabled by adopting electricity cooling system for a Ge detector. By attempting the miniaturization of the system, a compact field implementation plan is realized and also a periodic check can be simplified. (2) Thickness of a Ge detector is optimized to detect Xe-133 (80keV) as a main index nuclide to fuel leakage, so detection sensitivity is improved. (3) By adopting the mass flow control system wcich can control the flow stably in low, it is possible to reduce a volume and length of a decay tube for 13N which is the main interference nuclide, then a installation of the system is able to be simplified. (4) Moving average method which is effective to reduce noise is applied to a trend display of nuclides measured value, and it can be set at optimum time for the PST execution by flexibly setting a moving average time. *PST: Power Supression Test

The instrumentation and control (I&C) systems are important parts of the Nuclear Power Plants (NPPs). They are the key to safety, reliability and economic operation of the nuclear power generating units. With the development of systems at NPPs. Software reliability is important because software failure is usually considered as one form of common cause failures in digital I&C systems. The reliability analysis of the digital I&C systems, particularly quantitative evaluation on the software reliability analysis of the digital I&C systems belongs to a great challenge. This paper provides an overview of the ongoing research activities on a reliability analysis of digital I&C systems of NPPs, quantitative evaluation methods of software reliability, the advantages and disadvantages of the approaches, and research status at home and abroad. China is using the untraditional reliability analysis method of Multilevel Flow Model (MFM) to evaluate software reliability in digital I&C systems.

To improve the diagnosis accuracy and self-adaptability of steam turbine diagnosis systems in AP1000 NPP, this article attempts to apply a new method which unifies digital filter and stimulant filter into the filter processing of turbine Shaft vibration signal in Nuclear Power Plant. According to the process of power plant field data guideline, 4-level Butterworth filter prototype is used to resolve noise influence on original vibration signal. Supported by Matlab ANN Tools and Delphi, this study uses “spline” interpolation in the Process of orbit curve; with an ideal smooth effect achieved after compliance, this practice is proved that it can enhance the recognition effect of in Sanmen AP1000 Nuclear Power Plant TDM(Turbine Diagnosis Systems) monitoring functions.

With the application of nuclear technology in many fields and the development of nuclear power, the people pay more and more attention to which the radioactive substances are harmful to the human, environmental pollution and nuclear security. In order to prevent the dangers of radioactive substances on the professional staff and the public, the pollution of the environment, it is particularly important to establish a scientific and perfect radiation environment monitoring system. In this paper, the radiation environmental monitoring system based on network is designed for environment radiation situation of the original enterprise, such as the uranium mining, nuclear power plant, uranium hydrometallurgical plant, nuclear fuel processing plant. The network technology is introduced in the monitoring system which links the field monitoring system, data center service system, data transmission system and client system. the worker can control the on-site monitoring points, data acquisition and transmission in any of the network interface functions, complete displaying, analyzing and processing in the remote server, meanwhile the network power supply technology is applied in the monitoring system which can supply electricity to on-site monitoring points. According to the test, the radiation environmental monitoring system which based on network technology is easy to apply and maintain, with a certain degree of practicality and economy.

Nuclear power plant refueling cycle extension is an effective way of nuclear power plant to improve economic benefit. This paper, about the Qinshan Nuclear Power Plant phase II 1&2 unit, combination the practical refueling period experience of 12 months refueling period was extended to 18 months, focus on the instrument calibration extension methods of the refueling cycle extension project. Instrument in nuclear power plant running plays very important role, various instruments in the extension period after extended whether their properties still remain available and reliable, require detailed and reliable demonstration. Considering the Qinshan Phase II instrument equipments are purchased from multinational nations and different manufacturers, in this paper, the various demonstration methods are analyzed and compared, from all these methods, we chose the best one as the primary method in the actual demonstration, and use the others as supplemented methods by the demonstration strategies, and the main methods are detailed in the description and examples, so these can ensure that the instrument refueling cycle extension strategy is reliability.

Sanmen Nuclear Power Plant is constructing the first AP1000 unit in the world, and DCS system is adopted as the non-safety control system. This paper reviews the experience of Sanmen project, and analyzes the necessity of nuclear power plant Owner’s management for developing the DCS system, as well as how to manage the development effectively.

Now the entire safety I&C system is based on one kind of software and hardware platform, the Common Cause Failure (CCF) may impact the whole safety I&C system becomes to a potential risk. How to mitigate the effect of CCF in safety system and improve the safety of the nuclear power plant is considered by the system designer. Especially after the Fukushima nuclear accident, the Defense-in-Depth and Diversity (D3) should be more concerned by all designers. The diverse actuation system (DAS) plays a very important role in the D3 system. In this paper, the related codes and standards of DAS are analyzed firstly. Then, this paper expounds the approach to demonstrate the D3 analysis for the digital I&C systems applied to the nuclear power plant in detail. In the D3 analysis, all the safety functions of the digital safety system are assumed to be disabled by a CCF. DAS provides diverse automatic reactor trip and diverse safety injection actuation which are not impaired by the postulated CCF. DAS also provides manual actuation functions and plant parameter monitoring functions which can be used to cope with CCF. Finally, the paper takes the DAS of Fujian Fuqing Nuclear Power Plant as an example, introduce how to design the structure of the DAS and calculate the suitable setpoints.

A design of alarm system which includes both annuciator tiles and VDU alarm displays is proposed of the digital control room in HTR-PM (high temperature reactor pebble-bed module). Important and system-level alarms are located as fixed tiles, most of which are shared alarms. A shared alarm is a combination of several alarm causes associated with a single plant system or component. In the VDU displays, icons which stand for alarm tiles are developed. The tile icons obey the same coding rules of the real alarm tiles. A special VDU display is designed, in which the alarm list function and the tile icons are both included. The tile icons in the display are arranged exactly the same as the real tiles on the panel. When the shared alarm tile is activated, via the special VDU display, operators can easily get the detailed alarm information about the alarmed tiles with only a click on the tile icon. Based on the alarm display features, an alarm control procedure of pushbuttons on the console and the control functions in VDU display is presented. In the alarm response procedure, the personnel cannot acknowledge the alarm when the detailed messages are not read. This alarm system is technically based on the DCS (Distribute Control System) of the plant. Compared with using VDU alarm list only, tiles provide the operator perceptual understanding of the whole plant situation and rapid detection of alarms. The special VDU display with tile-like icons and function of listing detailed alarm messages provide the operator inquiry capability to obtain specific information, which support the operator effectively.