Preliminary definition of the remote handling system for the current IFMIF Test Facilities

A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.     

Contribution to DEMO reactor RH maintenance assessment

The scope of this paper is a preliminary assessment of the maintenance scheme in support of the European study for the next generation of fusion reactor: DEMO. Despite other fusion machine requiring remote handling maintenance operations, DEMO is supposed to work under steady state operational conditions. Therefore, requirement on the maintenance scheme is stronger. To target a good availability of the machine along machine operation plan, it is necessary to draw an adequate maintenance scheme. Indeed, due to the high fluxes generated by the plasma in the vacuum vessel, the first wall lifetime is limited, so the frequent replacement is necessary. On current fusion experimental machine, as first wall load conditions are less severe, DEMO condition implies high level of requirement on maintenance time. During DEMO lifetime, several full first wall replacements are expected. To provide access to the vacuum vessel machine for first wall removal, preparatory work is required to set the machine to adequate maintenance conditions and to open the machine properly, the same situation at the end of the maintenance period. Shutdown duration for first wall replacement should be as short as possible to reach the availability target of the machine. From this statement, the maintenance duration should not exceed 20% of the total lifetime of the reactor operation. First wall segmentation (i.e. total number of component to replace) has a high impact onto the replacement time. Considering the number of feasible designs for the first wall segmentation, we concentrate remote handling concept assessments one type of segmentation, the one minimizing the numbers of modules to replace [4], [5], [6]. Assumption on Divertor segmentation for these DEMO studies have similarities with Divertor ITER design; therefore ITER design output is relevant [1], [2]. We assume divertor removal performed in shadow time, while removing the other first wall modules.     

Design approach of the ARIES-AT power core and vacuum vessel cost assessment

The complexity of fusion power plants require the integration of many diverse and important system requirements to achieve a design approach that is viewed as a commercially viable electric plant. The ARIES-AT power core design builds upon a history of fusion power core designs that evolve along with physics and engineering advances. The baseline design point is optimized for maximum performance and minimum capital cost based upon the ARIES systems code results, along with physics and engineering analyses. The ARIES-AT power core is designed to be quick and easily maintainable to achieve high plant availability. A key element to achieve the high availability is the integration of the core elements with the design of the vacuum vessel. The vacuum vessel design is developed in more detail to assure the key assembly and maintenance features could be realized at an affordable cost.     

Design of a brazing connector for DEMO in-vessel components

The cost of electricity generated by fusion power will be strongly conditioned by the availability of future reactors. One key issue is the developing of feasible quick pipe connectors for the connection/disconnection of critical in-vessel components during maintenance operations. Brazing is a widely used joining technique which produces leak-proof high strength joints, with excellent stress distribution, little distortion and minimum oxidation. This work presents a design of a self-brazing/debrazing connector to be used with helium, lead–lithium and water pipes in DEMO. The remote handling compatible design includes an induction heating system, a brazing atmosphere supply, an inspection system (leak testing), a bolted/clamped union to provide stiffness against disruptions and thermal loads, and a positioning and alignment system.     

IoT-Based Maintenance Process Design for Fusion Reactor Remote Handling System

Nuclear fusion is one of the most important ways to resolve human energy shortages issues. Due to risk of ionizing radiation, the nuclear fusion reactor will relay on remote handling maintenance to achieve its scientific mission. Remote handling maintenance system is required to provide more reliable and powerful remote handling tools for increasing the efficiency of maintenance. With the development of technology, Internet of Things (IoT) has been became an important way to increase the maintenance efficiency. The basic structure of fusion machine and its functions has been introduced in this article. Besides, the maintenance system for fusion reactor has been described. Finally, IoT-based maintenance process for fusion reactor remote handling system has been established and it will greatly improve the maintenance efficiency and save the maintenance cost.     

Availability Analysis for CFETR

Availability analysis for Chinese fusion engineering testing reactor (CFETR) is very important at the pre-conceptual phase. Availability apportion are the theoretical basis of system design of CFETR. Availability analysis informs the development of the CFETR overall system and subsystem design. Availability analysis will enable the identification of key subsystems to achieve availability targets. The duty cycle of CFETR should be at least 0.3–0.5. Such design goals require all subsystems of CFETR must have a pretty high availability. The availability of CFETR can be defined by break time analysis results. Analysis results proved that the availability of CFETR is 0.5–0.7. All availability subsystems of CFETR can be apportioned from their relevant mean time before failure (MTBF) and mean time to repair (MTTR) data to meet the availability goals. The relation between reliability, maintainability and availability indicates that a subsystem could have a high availability though its reliability is pretty low. Availability apportionment and analysis indicate that the availability of blanket and divertor should be 0.769–0.91 to meet the design requirements and their availability can be improved by increase the MTBF and reduce the MTTR of blanket and divertor with the development of remote handling or remote maintenance technology for fusion reactor.     

ARIES-AT maintenance system definition and analysis

A fusion power plant must have a high availability to be competitive in the electrical generation market. Attaining high plant availability is difficult because the fusion power core has a limited service lifetime. Moreover, the core components are radioactive and very large. To assess these issues, the maintainability of the ARIES fusion power core is analyzed and integrated into the early power core design process, which results in a maintainability approach capable of attaining a relatively short refurbishment time. The developed timelines are presented for the scheduled maintenance of the power core. The short core refurbishment time coupled, with evolutionary improvements in the maintainability of the reactor plant equipment and the balance-of-plant equipment, infer an attractive plant availability in the range of 90%.     

Author index to volume 63

STARFIRE is a design for a conceptual commercial tokamak electrical power plant based on the deuterium/tritium/lithium fuel cycle. In addition to the goal of being technologically credible, the design incorporates safety and environmental considerations. STARFIRE is considered to be the tenth in a series of commercial fusion power plants.STARFIRE has a 7-m major radius reactor producing 1200 MW of net electrical power from 4000 MW of thermal power, with an average neutron wall load of 3.6 MW/m². The aspect ratio is 3.6 and a D-shaped plasma with a height-to-width ratio of 1.6 and average toroidal beta of 0.067 is used. The maximum magnetic field is 11T. Availability goals have been set at 85% for the reactor and 75% for the complete plant including the reactor.The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum for impurity control, most superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield.     

STARFIRE,a commercial tokamak power plant design

STARFIRE is a design for a conceptual commercial tokamak electrical power plant based on the deuterium/tritium/lithium fuel cycle. In addition to the goal of being technologically credible, the design incorporates safety and environmental considerations. STARFIRE is considered to be the tenth in a series of commercial fusion power plants.STARFIRE has a 7-m major radius reactor producing 1200 MW of net electrical power from 4000 MW of thermal power, with an average neutron wall load of 3.6 MW/m². The aspect ratio is 3.6 and a D-shaped plasma with a height-to-width ratio of 1.6 and average toroidal beta of 0.067 is used. The maximum magnetic field is 11T. Availability goals have been set at 85% for the reactor and 75% for the complete plant including the reactor.The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum for impurity control, most superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield.     

Pre-conceptual design assessment of DEMO remote maintenance

EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.     

Efficiency and availability driven R&D issues for DEMO

DEMO is the main step foreseen after ITER to demonstrate the technological and commercial viability of a fusion power plant. DEMO R&D requirements are usually identified on the basis of the functions expected from each individual system. An approach based on the analysis of overall plant functional requirements sheds new light on R&D needs. The analysis presented here focuses on two overall functional requirements, efficiency and availability. The results of this analysis are presented here putting emphasis on systems not sufficiently considered up to now, e.g. the heating and current drive systems, while more commonly addressed systems such as tritium breeding blankets are not discussed in detail. It is also concluded that an overall functional analysis should be adopted very early in the DEMO conceptual design studies in order to provide a fully integrated approach, which is an absolute requirement to ensure that the ambitious goals of this device will be ultimately met.     

Proposal of an improved design of IFMIF Test Cell components for enhanced handling and reliability

An adequate design of components to be manipulated by remote handling is a key factor in the success of any activated facility, having a decisive impact on availability, prompt and safe maintenance, occupational exposures and flexibility of the facility. Such components should satisfy at least the basic remote handling requirements of simplicity, accessibility, modularity, standardization and assembling adequacy. Highly activated components in the IFMIF facility are found in the Test Cell, a pit closed by stepped shielding plugs. The Test Cell confines the Test Modules which contain the samples and experiments. The present reference design of the IFMIF Test Cell shows some drawbacks, in particular the jamming tendency of the shielding plugs, slow and complex access to the Backplate, a low lifetime component, and difficult positioning of the Test Modules. This paper summarises several modifications aiming at improving, under such remote handling requirements, the present reference design of the Test Cell shielding plugs and aspects of the geometrical structure of the Test Cell. A functional modularization of the present shielding plugs has been carried out and positioning guides for the Test Modules have been devised.     

RAMI analysis of the ITER Central Safety System

ITER is the first worldwide international project aiming to design a facility to produce nuclear fusion energy. The technical requirements of its plant systems have been established in the ITER Project Baseline. In the project, the Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to help aid the design of the components in preparation for operation and maintenance. A RAMI analysis was performed on the conceptual design of the ITER Central Safety System (CSS). A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 2 main functions and 20 sub-functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CSS expected after implementation of mitigation actions was calculated to be 99.80% over 2 years, which is the typical interval of the scheduled maintenance cycles. This is consistent with the project required value of 99.9 ± 0.1%. A Failure Modes, Effects and Criticality Analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability of the plasma operation. This analysis defined when risk mitigation actions were required in terms of design, testing, operation procedures and/or maintenance to reduce the risk levels and increase the availability of the main functions.     

Waste management scenario in the hot cell and waste storage for DEMO

Management strategy of radioactive waste generated in periodic replacement may be important in the point of view of fusion reactor design, because it has a large impact on the design of the hot cell and waste storage in the plant. In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules needs to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain reasonable plant availability. In the hot cell, the modules will be removed from the back plate of the assembly. Here, note that the active cooling must be done by a way that does not cause contamination of the hot cell environment due to dispersion of tritium and tungsten dust. In this sense, the cooling scenario is adopted that the existing pipe of cooling water in the assembly is connected to a different cooling water system in the hot cell. On the other hand, it is assumed that the structural material (F82H) of the blanket and divertor is not recycled due to its high contact dose rate. It should be crushed into small pieces to reduce volume of the waste and required storage space. In this paper, the basic idea of the waste management scenario and the conceptual design in the hot cell and waste storage for DEMO has been proposed.     

核电站设备可靠性数据库的建立与应用

以大亚湾核电站为例,论述了核电站设备可靠性数据的采集与处理、可靠性参数的分析计算方法及可靠性数据库的建立与应用等。数据源涉及到设备的设计信息、运行信息、维修信息、定期试验记录等。在大亚湾核电站运行经验的基础上,形成了大亚湾核电站设备可靠性数据库,为深入地、客观地记录核电站各类设备的运行历史和现状、监控电站设备,特别是与安全相关设备的状态,提供了有效的工具;为加强设备的可靠性与可用性管理、确保电站的安全经济运行,提供了非常有实用价值的信息;同时还为核电站的安全管理、可靠性分析、概率安全评价、以可靠性为中心的维修及经济性管理等领域里的新技术在核电站的应用研究与开发,提供了必不可少的数据。     

LIBRA-LiTE: A 1000 MWe reactor

Conclusions The results from this study indicate that light ions can be a competitive factor in the race to commercial fusion power. The relatively simple and near-term driver technology is particularly attractive compared to higher cost laser and heavy ion schemes. The cavity design and engineering operations can be tailored such that Utilities could envision a reliable and maintainable power plant. The major problem to be faced now is the method of beam propagation to the target. The LIBRA-LiTE design reveals that ballistic transport may be more attractive from a physics standpoint, but the severe neutron environment presents a challenge to materials scientists. Continued experimentation and research is needed to develop a truly attractive ICF power plant.     

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.     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.