ITER Cryostat—An overview and design progress

The ITER Cryostat is one of the most important and critical systems in the ITER project. It envelops the entire basic systems of the Tokamak and is a vacuum tight container. Cryostat provides vacuum environment for the thermal insulation to magnet system operating at 4.5k and thermal shield system operating at 80k. It is evacuated to a pressure of 10^?4 Pa prior to cool down of the magnets and thermal shields in order to limit heat transfer by gas conduction and convection to a level tolerable to the cryogenically cooled components. The cryostat is also designed to support all the loads like gravity, electromagnetic forces, seismic, etc. that derive from the Tokamak basic systems, and from the Cryostat itself, to the floor of the Tokamak pit through its support structures during the normal and off-normal operational regimes, and at specified accidental conditions. ITER Cryostat conceptual design was reviewed last year and detailed design is reviewed in June 2010. At present Cryostat design is in its final stage of completion and final design review is planned in October 2010 to finalize the procurement specifications. Procurement Arrangement will be signed this year with Indian Domestic Agency and procurement cycle will start for the fabrication of the ITER Cryostat. This paper discusses the updated Cryostat design and analysis with integration of the different penetrations required for the communication within the In-Cryostat system and its maintenance.     

ITER remote maintenance system configuration model overview

A challenge for the ITER project is to manage the design of many systems being developed in parallel. In order to control the machine configuration and ensure proper design integration, the ITER project has implemented the so-called “configuration management models” (CMMs), aimed at controlling and managing the machine systems’ interfaces. Specific issues are raised for modelling the ITER remote maintenance system (IRMS). That system shall provide the mean to support the remote maintenance operations for in-vessel components, remote transfer of activated components between the vacuum vessel (VV) and the hot cell facility and remote repairing, refurbishing and/or processing operations in the hot cell facility.The IRMS are dynamic, constantly changing morphologies, working envelopes and locations within the plant. This raises the issue of how to integrate the dynamic nature of this equipment into the CMM required for design integration. This paper describes the design methodology that is being developed to address the specific nature of the IRMS in the building of the CMM and gives examples to demonstrate the benefits to be gained by adopting this approach.     

Magnetic compatibility of standard components for electrical installations: Computation of the background field and consequences on the design of the electrical distribution boards and control boards for the ITER Tokamak building

Inside the proposed Tokamak building, the ITER poloidal field magnet system would produce a stray magnetic field up to 70 mT. This is a very unusual environmental condition for electrical installation equipment and limited information is available on the magnetic compatibility of standard components for electrical distribution boards and control boards. Because this information is a necessary input for the design of the electrical installation inside the proposed ITER Tokamak building specific investigations have been carried out by the ITER European Participant Team. The paper reports on the computation of the background magnetic field map inside the ITER Tokamak building and the consequences on the design of the electrical installations of this building. The effects of the steel inside the building structure and the feasibility of magnetic shields for electrical distribution boards and control boards are also reported in the paper. The results of the test campaigns on the magnetic field compatibility of standard components for electrical distribution boards and control boards are reported in companion papers published in these proceedings.     

Magnetic fusion commercial power plants

Toroidal magnetic systems offer the best opportunity to make a commercial fusion power plant. They have, between them, all the features needed; however, no one system yet meets the ideal requirements. The tokamak is the most advanced system, and the proposed International Thermonuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX) will build upon the existing program to prepare for an advanced tokamak demonstration plant. Complementary toroidal systems such as the spherical torus, stellarator, reversed-field pinch, field-reversed configuration, and spheromak offer, between them, potential advantages in each area and should be studied in a balanced fusion development program.     

Design modification and optimization of the ITER cooling water system

ITER (Latin for “the way”), the largest fusion experimental reactor in the world, is designed to demonstrate the technological feasibility of nuclear fusion energy conversion, at plant scale, from high temperature deuterium-tritium plasma using the Tokamak magnetic confinement arrangement.ITER will have a large vacuum vessel that hosts the plasma facing components. These components include the blanket and the divertor that will operate at temperatures, heat loads, and neutron flux higher than those reached in a nuclear fission power plant reactor.One of the main critical issues of the ITER reactor is the design of the cooling water system to transfer the heat generated in the plasma to the in-vessel components and the heat loads from the auxiliary systems to the environment.This paper describes the current ITER cooling water system and recent design modifications and optimizations.     

Design and technical status of the EU contribution to ITER

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R&D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost.This paper provides an update of the design and technical status of EU contributions to ITER.     

Progress in the design of ITER front-end vacuum instrumentation and controls

The ITER vacuum system will be one of the largest and most complex vacuum systems ever to be built. Extensive instrumentation and controls are being developed to satisfy the stringent vacuum processes necessary for the successful and safe operation of the ITER Tokamak. The complexity and deep integration of the vacuum systems within the ITER machine presents a challenge to implement all of the controls necessary for reliable operation. Several thousand valves and sensors have to be implemented within the harsh environmental conditions of the Tokamak vicinity, and require engineering of instrumentation and controls with remote electronics solutions.In this paper the status of the design of field end vacuum controls and instrumentation for the ITER vacuum systems is described. Details of the progress on selection of sensors and actuator technologies are given herein and solutions for remote device operation, including those for cryogenic devices, are described together with necessary local shielding.     

Optimization of the electrostatic and magnetic field configuration in the MITICA accelerator

MITICA (Megavolt ITER Injector Concept Advancement) is a test facility for the development of a full-size heating and current drive neutral beam injectors for the ITER Tokamak reactor. The optimized electrostatic and magnetic configuration has been defined by means of an iterative optimization involving all the physics and the engineering aspects. The acceleration grids have been designed considering optical performances and mechanical constraints related to embedded magnets, to cooling channels, to the grid stiffness and manufacturability. A combination of “local” vertical field and horizontal “long range” field has been found to be the most effective set-up for ion extraction, beam focusing and minimization and equalization of thermo-mechanical loads and minimal number of electrons exiting the accelerator.     

Preliminary design of the control and data acquisition systems for the Neutral Beam Test Facility

The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600 s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC.     

Seismic design of the ITER main Tokamak components

In order to assess the structural performance of the ITER Main Components it is important to take into account not only the mutual dynamic interaction among them during a seismic event but also their interactions with the Tokamak Buildings (TB) complex. The seismic behavior of the TB is affected by the large dimensions of the building, the concrete basemat thickness that has to be sufficiently rigid to support the weight of the Tokamak, the presence of anti-seismic bearing (ASB) under the basemat, and the distribution of heavy equipment at higher levels. These factors require that the soil–structural interaction must be studied in detail, taking into account the specific effects such as the excavation influence and the building rocking motion due to seismic wave propagation. The study of the seismic behavior has been carried out using two different linear dynamic methodologies: power spectral density (PSD) and spectral analyses. The paper illustrates the main results of the seismic analyses and gives the seismic design input for the Tokamak components in terms of support loads, accelerations and displacements.     

Updated safety analysis of ITER

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.     

基于EPICS的J-TEXT CODAC系统

J-TEXT装置是华中科技大学恢复建造的中型托卡马克装置,已于2007年放电运行,其控制系统采用分布式结构,由多个子系统组成。为提高子系统集成、维护和更新的效率,并有效地管理各子系统、控制装置的运行状态及保障设备和人员安全,J-TEXT装置参考ITER CODAC的设计思路,结合J-TEXT装置的需求设计了J-TEXT CODAC系统。J-TEXT CODAC系统为装置各子系统提供统一的设计模型和相关设计标准,使用EPICS软件作为通讯中间层,设计了全局控制系统、时序和同步控制系统、联锁保护系统,并将原有控制系统改造、集成到J-TEXT CODAC系统中。目前该系统已部署在J-TEXT装置上,在2012年春季以来的多轮实验中运行良好。     

Application of GEMSAFE to ITER CDA and Its comparison with FER

A safety analysis for the design of International Thermonuclear Experimental Reactor (ITER) in the Conceptual Design Activity stage was performed by the GEMSAFE methodology, and its results were compared with those of Fusion Experimental Reactor (FER), a Japan's facility planned next to JT-60. The objectives of this study are to confirm the applicability of GEMSAFE to ITER and to select design basis events of ITER and identify R&D items with comparison to FER. Function-Based Safety Analyses (FBSA) were carred out to select 19 and 25 design basis events for FER and ITER, respectively. The major reason for the difference is that ITER has a class-2 RI source, e.g., tritium of 7.5 × 10⁵ Ci in mobile form, in the coolant for the first wall and blankets as well as a class-3 RI source, e.g., the immobile tritium of 2.2×10⁷ Ci absorbed in first wall and dust.     

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.     

Design progress on ITER port plug test facility

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment.The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs.Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 × 10^?5 Pa at 100 °C containing a port plug. The heating system shall provide water at 240 °C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests.This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.     

HL-2A托卡马克装置的工程和实验概况

中国环流器二号A装置(HL-2A)是核工业西南物理研究院2002年投入实验运行的托卡马克,它是我国第1个具有偏滤器、等离子体截面具有一定垂直拉长的托卡马克.HL-2A的磁体使用铜导体,具有良好的灵活性和等离子体的可近性,其极向场线圈全部位于环向场线圈之内,位于真空室内的偏滤器的成形线圈可建立双零和单零的偏滤器位形.HL-2A已发展了30多套先进的等离子体诊断系统和总功率4 MW的辅助加热系统,加料技术得到持续发展.随着上述系统的建设和放电综合控制技术的提高,HL-2A装置已获得了高约束模式,这为开展先进托卡马克(AT)物理实验,ITER和聚变堆的科学、技术和工程问题等的研究奠定了基础.HL-2A也成为国际上最活跃的中型托卡马克,为国际托卡马克物理活动(ITPA)作出了积极贡献.     

Preliminary Occupational Radiation Exposure evaluation related to NET/ITER tritium systems

This paper presents the criteria adopted to evaluate Occupational Radiation Exposure (ORE) during normal operation and maintenance of NET/ITER and some results concerning the fuel cycle systems located in the tokamak and tritium buildings. Prompt radiation, activity concentration, and intake situations as well as number of workers, number of events, and exposure time are considered. Many systems and components, whose location in the plant can affect radiological protection during maintenance and/or surveillance, are identified together with the operations needed for each activity. Accidental conditions and equipment failures have been considered in the special maintenance activity when they are due to events with a high probability of occurrence so that such events might be expected during the life of the plant. Some results are reported showing the ORE figures with reference to the main activities. The total man-Sv/y for the systems and activities considered is about 0.5. Such a result, even if very preliminary and incomplete, means that ORE for the tritium systems of a machine like NET/ITER is not negligible and has to be continuously controlled during the design phase.     

Status of ITER dimensional tolerance studies

The optimization of the manufacturing/assembly tolerances and processes in ITER Experimental Nuclear Fusion Device is one of the key tasks to optimize the fabrication cost, to prevent problems during assembly and to ensure that the critical homogeneity of the magnetic field and the positioning requirements of the plasma facing components can be achieved. This task is further complicated by the strong interplay among the various Tokamak systems, as for instance in the inner region of the machine where the clearances between Central Solenoid, Toroidal Field Coils, Thermal Shield, Vacuum Vessel and In-Vessel components have been minimized for their large influence on the magnetic flux and the overall machine cost.A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. The dimensional variation model is representative of Tokamak functional tolerances and processes, predicting accurate values for the amount of variation on critical areas. In addition, dimensional simulations help to determine the key tolerances that contribute to a particular variation.This paper describes the current status of the Tokamak dimensional variation studies and its management plan, highlighting the status of compliance of allocated tolerances with input requirements. Management of risk issues and corrective actions are also described.     

The ITER divertor pumping system,design evolution,simplification and performance

The ITER tokamak will be fuelled at a time averaged rate of up to 200 Pam³ s^?1 requiring neutralised gas in the divertor to be pumped to balance the fuelling and remove the fusion helium and other impurities in the exhaust. This is achieved on ITER using large bespoke cryo-sorption pumps. In this paper design evolution of the ITER divertor pumping system is outlined from the 1998 configuration to the current design. Details of the new, 6 direct pump, system design which will be used in the build of ITER are given. The operating modes of the new system configuration for different plasma scenarios are described and the performance of the new system is analysed and compared with previous baselines.     

Analysis of ripple and error field due to using the ferritic steel insert in EAST

Chinese Experimental Advanced Superconducting Tokamak (EAST) is ITER-like Superconducting (SC) Tokamak with divertor configuration. However, EAST device has 16 toroidal field coils (TFCs) whose ripple amplitude is 0.67% higher than 0.3% of acceptable level of ITER at separatrix point. In order to improve the plasma control and confinement and have more contribution to ITER Physics, it is expected to reduce the TF ripple to ITER acceptable level. In this contribution, it was preliminarily investigated for installation of the appropriate ferristic steel insert inside EAST vacuum vessel to reducing the ripple based on electromagnetic analyses. Results indicated the ripple amplitude could be achieved to the expected level of less than 0.3%. Simultaneously, the error fields introduced due to installation of the ferristic steel insert was analyzed and not beyond scope of physics requirement.