Main control system verification and validation of NPP digital I&C system based on engineering simulator

Full-scope digital instrumentation and controls system (I&C) technique is being introduced in Chinese new constructed Nuclear Power Plant (NPP), which mainly includes three parts: control system, reactor protection system and engineered safety feature actuation system. For example, SIEMENS TELEPERM XP and XS distributed control system (DCS) have been used in Ling Ao Phase II NPP, which is located in Guangdong province, China. This is the first NPP project in China that Chinese engineers are fully responsible for all the configuration of actual analog and logic diagram, although experience in NPP full-scope digital I&C is very limited. For the safety, it has to be made sure that configuration is right and control functions can be accomplished before the phase of real plant testing on reactor. Therefore, primary verification and validation (V&V) of I&C needs to be carried out. Except the common and basic way, i.e. checking the diagram configuration one by one according to original design, NPP engineering simulator is applied as another effective approach of V&V. For this purpose, a virtual NPP thermal-hydraulic model is established as a basis according to Ling Ao Phase II NPP design, and the NPP simulation tools can provide plant operation parameters to DCS, accept control signal from I&C and give response. During the test, one set of data acquisition equipments are used to build a connection between the engineering simulator (software) and SIEMENS DCS I/O cabinet (hardware). In this emulation, original diagram configuration in DCS and field hardware structures are kept unchanged. In this way, firstly judging whether there are some problems by observing the input and output of DCS without knowing the internal configuration. Then secondly, problems can be found and corrected by understanding and checking the exact and complex configuration in detail. At last, the correctness and functionality of the control system are verified. This method is also very convenient for expansion to other type digital I&C V&V. This paper is mainly focused on V&V of closed-loop control systems in full-scope DCS and several detailed reactor control (RRC) systems, including pressurizer pressure and water level control, steam generator water level control. The V&V works were carried out by applying engineering simulator. This paper describes the structure and function of the simulator, V&V procedure, results analysis and problems identified. Through the actual on-line virtual closed-loop testing on Ling Ao Phase II NPP project, many problems of DCS configuration were found and solved. And it proved that V&V based on engineering simulator enables significant time saving, improves economics and safety in the phase of engineering debugging.     

The first digital reactor protection system in China

The reactor protection system (RPS) used in the 10 MW high-temperature gas-cooled reactor is the first digital RPS designed and operated in China. In order to ensure its safety and reliability and to reduce the development risk and cost, some measures had to be taken. The measures adopted in the development process include the architecture of defense-in-depth, commercial grade hardware, prototype development model, separation of safety class software and non-safety class software, deterministic behavior of safety software, etc. The measures adopted in the verification and validation (V&V) process include effective dedication on the commercial grade hardware, emphasis on the assessment of the requirements and specifications, emphasis on the demonstration and testing, thorough testing for the safety function, long period demonstration operation, application of automatic test system to improve the efficiency of V&V processes, etc. As a result, this first digital RPS has passed the safety assessment of the National Nuclear Safety Authority. Its performance and safety are proven to be confident and assuring through the demonstration and testing. Thus, the design and V&V process of the first digital protection system in China was successful.     

Technological review of the HRP manufacturing process R&D activity

ENEA and Ansaldo Nucleare S.p.A. have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities for the manufacturing of high heat flux plasma-facing components (HHFC), and in particular for the inner vertical target (IVT) of the ITER divertor.This component has to be manufactured by using both armour and structural materials whose properties are defined by ITER. Their physical properties prevent the use of standard joining techniques. The reference armour materials are tungsten and carbon/carbon fibre composite (CFC). The cooling pipe is made of copper alloy (CuCrZr-IG).During the last years ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components of different length, geometry and materials, by using innovative processes: HRP (hot radial pressing) and PBC (pre-brazed casting).The history of the technical issues solved during the R&D phase and the improvements implemented to the assembling tools and equipments are reviewed in the paper together with the testing results.The optimization of the processes started from the successful manufacturing of both W and CFC armoured small scale mockups thermal fatigue tested in the worst ITER operating condition (20 MW/m²) through the achievement of record performances obtained from a monoblock medium scale mockup.On the base of these results ENEA-ANSALDO participated to the European programme for the qualification of the manufacturing technology to be used for the procurement of the ITER divertor IVT, according to the F4E specifications. A divertor inner vertical target prototype (400 mm total length) with three plasma facing component units, was successfully tested at ITER relevant thermal heat fluxes.Now, ANSALDO and ENEA are ready to face the challenge of the ITER inner vertical target production, transferring to an industrial production line the experience gained in the development, optimization and qualification of the PBC and HRP processes.     

HTR-10数字化保护系统的验证与确认

10MW高温气冷实验堆(HTR-10)验证与确认(VerificationandValidation,简称V&V)过程的特点是:和开发过程紧密配合,特别重视设计说明书的V&V、安全软件的V&V、系统功能的完整测试等环节。从而使数字化保护系统成功地通过安全审查并被批准应用于HTR-10上。     

The European contribution to the ITER Remote Maintenance

For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source.The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R&D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages.Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations.After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems.Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.     

Manufacturing preparations for the European Vacuum Vessel Sector for ITER

The contract for the seven European Sectors of the ITER Vacuum Vessel, which has very tight tolerances and high density of welding, was placed at the end of 2010 with AMW, a consortium of three companies. The start-up of the engineering, including R&D, design and analysis activities of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. The statutory and regulatory requirements of ITER Organization and the French Nuclear Safety regulations have made the design development subject to rigorous controls. AMW was able to make use of the previous extensive R&D and prototype work carried out during the past 9 years, especially in relation to advanced welding and inspection techniques. The paper describes the manufacturing methodology with the focus on controlling distortion with predictions by analysis, avoiding use of welded-on jigs, and making use of low heat input narrow-gap welding with electron beam welding as far as possible and narrow-gap TIG when not. Further R&D and more than ten significant mock-ups are described. All these preparations will help to assure the successful manufacture of this critical path item of ITER.     

ITER articulated inspection arm (AIA): R&d progress on vacuum and temperature technology for remote handling

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R&D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER.It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs.To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 °C).The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility.As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system.This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.     

Results of CMM standalone tests at DTP2

A full scale physical test facility, DTP2 (Divertor Test Platform 2) has been established in Finland for demonstrating and refining the Remote Handling (RH) equipment designs for ITER. The first prototype RH equipment at DTP2 is the Cassette Multifunctional Mover (CMM) equipped with the Second Cassette End Effector (SCEE) delivered to DTP2 in October 2008. The purpose is to prove that CMM/SCEE prototype can be used successfully for the 2nd cassette RH operations. At the end of F4E grant “DTP2 test facility operation and upgrade preparation”, the RH operations of the 2nd cassette were successfully demonstrated to the representatives of Fusion for Energy (F4E).During the grant the High Level Control (HLC) software developed at DTP2 was integrated with the CMM/SCEE hardware. The performance criteria of the CMM/SCEE equipment were defined based on the ‘EN ISO 9283 Manipulating industrial robots – Performance criteria and related test methods’ standard. Considerable improvement to the performance was achieved with the aid of compensation functions, which took into account the deflections and the compliance effects caused by the Divertor Cassette weighting 9000 kg. According to measurements the positioning error at the furthest point of the cassette was reduced from 80 mm to 5 mm.So far the 2nd cassette mock-up has been installed and removed already some tens of times. The reliability of the HLC software is sufficient to operate the CMM/SCEE all day without interruptions. Also the execution of the automatic RH processes with the overall RH system is reliable and repeatable in terms of accuracy and cycle time.These experiments provide a solid basis for investigating the RH system ability to perform fail-safe operations in various failure scenarios and to recover from them. The target of the continuing R&D is to find out a more complete set of functional and non-functional requirements for the RH system for Divertor Cassette maintenance to ensure an adequate level of requirements and procedures for ITER.     

Overview of the design and R&D of the ITER blanket system

The ITER blanket design has substantially evolved since the ITER design review of 2007. Two major incentives for the design changes have been the need to account for large plasma heat fluxes to the First Wall (FW) and the need for acceptable maintenance of FW panels. In parallel to the design effort, a focused R&D program is being carried out including manufacturing and testing of semi-prototypes for the FW panels, and of full-scale prototypes for the shield blocks. This paper summarizes the status of the ITER blanket system design including the accommodation of interfaces, and describes some of the key R&D activities in support of the design with the goal of starting procurement in the first half of 2013.     

ITER Fast Plant System Controller prototype based on PXIe platform

The ITER Fast Plant System Controller (FPSC) is based on embedded technologies. The FPSC will be devoted to both data acquisition tasks (sampling rates higher than 1 kHz) and control purposes (feedback loop actuators). Some of the essential requirements of these systems are: (a) data acquisition and data preprocessing; (b) interfacing with different networks and high speed links (Plant Operation Network, timing network based on IEEE1588, synchronous data transference and streaming/archiving networks); and (c) system setup and operation using EPICS (Experimental Physics and Industrial Control System) process variables.CIEMAT and UPM have implemented a prototype of FPSC using a PXIe (PCI eXtension for Instrumentation) form factor in a R&D project developed in two phases. The paper presents the main features of the two prototypes developed that have been named alpha and beta. The former was implemented using LabVIEW development tools as it was focused on modeling the FPSC software modules, using the graphical features of LabVIEW applications, and measuring the basic performance in the system. The alpha version prototype implements data acquisition with time-stamping, EPICS monitoring using waveform process variables (PVs), and archiving. The beta version prototype is a complete IOC implemented using EPICS with different software functional blocks. These functional blocks are integrated and managed using an ASYN driver solution and provide the basic functionalities required by ITER FPSC such as data acquisition, data archiving, data pre-processing (using both CPU and GPU) and streaming.     

Gripping tool for the ITER upper port plug RH extraction/insertion

The conceptual design of several gripping tools and their mechanical interfaces is being carried out for the ITER ECH UPP within the WP10-GOTRH programme. EFDA finances the GOT RH (Goal Oriented Training Programme for Remote Handling). The purpose of this paper is to introduce new concepts of gripping tools for the plug extraction/insertion in the upper port of ITER. All these gripping tools are designed according to IO input data and geometrical constraints. The gripping tools have to be able to extract/insert the plug in the scenario of maximum misalignment between the plug and the tractor. The paper also defines the functional requirements the gripping tools need to comply with. The requirements and input data are verified and validated through 3D simulation with Catia mock-ups of the gripping tools. The strengths and weaknesses of each gripping tool model are compared.     

Software fault detection and recovery in critical real-time systems: An approach based on loose coupling

Remote handling (RH) systems are used to inspect, make changes to, and maintain components in the ITER machine and as such are an example of mission-critical system. Failure in a critical system may cause damage, significant financial losses and loss of experiment runtime, making dependability one of their most important properties. However, even if the software for RH control systems has been developed using best practices, the system might still fail due to undetected faults (bugs), hardware failures, etc. Critical systems therefore need capability to tolerate faults and resume operation after their occurrence. However, design of effective fault detection and recovery mechanisms poses a challenge due to timeliness requirements, growth in scale, and complex interactions. In this paper we evaluate effectiveness of service-oriented architectural approach to fault tolerance in mission-critical real-time systems. We use a prototype implementation for service management with an experimental RH control system and industrial manipulator. The fault tolerance is based on using the high level of decoupling between services to recover from transient faults by service restarts. In case the recovery process is not successful, the system can still be used if the fault was not in a critical software module.     

On design and development of additional End-Effectors for the Cassette Multifunctional Mover

The divertor area of ITER Vacuum Vessel (VV) consists of 54 modular cassettes which must be replaced three times during the estimated 20 years of operation of the ITER. Cassette Multifunctional Mover (CMM) and Cassette Toroidal Mover (CTM) are used in the cassette remote handling (RH).In this paper we discuss the design and development process for the RH equipment to be used in the ITER environment. Design concepts for the Standard Cassette End-Effector and Central Cassette End-Effector are described and the conceptual design phase methodology is presented.The main improvements of the new End-Effector concept designs are more robust and reliable assembly process with reduced CMM mover assembly accuracy requirement. New Central Cassette locking system was developed to address the high forces and contact pressures emerging during the Central Cassette installation. The chosen design concepts are verified with virtual reality simulations and are fulfilling the requirements defined in the concept design phase, including structural, assembly sequence, safety and reliability.     

Verification and validation of the HTGR systems CFD code Flownex

Regulatory requirements prescribe extensive verification and validation (V&V) of computer codes that are used in the design and analysis of accident conditions in nuclear plants. Flownex is a dynamic systems CFD code used as the primary thermal-fluid simulation code by the Pebble Bed Modular Reactor Company (PBMR).Stringent quality assurance processes have been implemented to ensure that the code conforms to the set standards. These processes include the comparison of Flownex with analytical results as well as with experimental data.The results of this process are summarized in this paper. Analytical solutions are used to verify Flownex's element models so as to ensure that the basic theory is correctly implemented in the computer code. As part of the analytical V&V effort various well-defined problems are solved using numerical methods implemented in independent computer codes.Comparison with experimental and plant data is a very important feature of the V&V program to validate that the chosen theory is fit for purpose. For this, validation data from the pebble bed micro model (PBMM) is used. In addition to the PBMM experimental data Flownex is compared to a number of small thermal-fluid experiments in which certain specific component phenomena is validated. These experiments were developed in collaboration with North-West University (previously Potchefstroom University).     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

The design of the ITER first wall panels

The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R&D activities.     

Developments in remote metrology at JET

The need to maximise the operational availability of fusion devices has driven the enhancements in accuracy, flexibility and speed associated with the inspection techniques used at JET. To this end, the remote installation of the ITER-Like Wall (ILW) tiles, conduits and embedded diagnostics has necessitated the adoption of technologies from other industries for their use in conjunction with the JET Remote Handling (RH) system. The novel adaptation of targetless stereophotogrammetry, targeted single-camera photogrammetry and gap measurement techniques for remote applications has prompted a range of challenges and lessons learnt both from the design process and operational experience.Interfacing Commercial Off-The-Shelf (COTS) components with the existing RH equipment has highlighted several issues of relevance to the developing ITER RH system. This paper reports results from the stereophotogrammetry and the single-camera photogrammetry surveys, allowing analysis of the effectiveness of the RH system as a platform for in-vessel measurement. This includes scrutiny of the accuracy achieved with each technique as well as the impact on the in-vessel Configuration Management Model (CMM). The paper concludes with a summary of key recommendations for the ITER RH system based on the experience of remote metrology at JET.     

ITER重力支撑的制造设计、认证测试及关键技术研究

重力支撑（GS）作为国际热核聚变实验堆（ITER）磁体支撑系统的关键部件,不但要承受环向场超导磁体净重以及交变的电磁载荷,同时还需隔离来自杜瓦环的热量以维持环向场超导线圈的热稳定性。本文通过有限元分析和工程测试验证了GS结构设计的可靠性;通过换热分析和真空热交换效率测试验证了热锚连接结构的可靠性;通过全尺寸螺栓77 K疲劳测试验证了螺栓原型件的疲劳性能。在随后的制造过程中,使用液压拉伸器和研制的高精度螺栓伸长量测量装置对所有的螺栓进行了均匀、精确地紧固。真空正压氦检漏的测试结果证明了GS的泄漏率远低于ITER的要求。基于以上工程测试的结果,本文设计的GS的结构是可行的且能运用于ITER装置中。