Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined.US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.     

Progress in the design,R&D and procurement preparation of the ITER Divertor Remote Handling System

The ITER Divertor Remote Handling System (DRHS) consists of a number of dedicated remote handling equipment and tooling that will provide the means to perform the exchange of the divertor system in a full-remote way. In order to achieve this objective the DRHS will need to perform a number of novel and complex remote operations in a contaminated and space-constrained environment, in rather poor lightening conditions. Fusion for Energy has recently launched the tendering phase for the in-kind procurement of the DRHS. The procurement is based on a set of system requirements and functional specifications supported by a reference design which are presented and discussed in this paper along with the main outcomes of the different R&D activities that have contributed to the development and validation of the current reference design.     

Development of the bus joint for the ITER Central Solenoid

The terminations of the Central Solenoid (CS) modules are connected to the bus extensions by joints located outside the CS in the gap between the CS and torodial field (TF) assemblies. These joints have very strict space limitations. Low resistance is a common requirement for all ITER joints. In addition, the CS bus joints will experience and must be designed to withstand significant variation in the magnetic field of several tenths of a Tesla per second during initiation of plasma. The joint resistance is specified to be less than 4 nΩ.The joints also have to be soldered in the field and designed with the possibility to be installed and dismantled in order to allow cold testing in the cold test facility. We have developed coaxial joints that meet these requirements and have demonstrated the feasibility to fabricate and assemble them in the vertical configuration. We introduced a coupling cylinder with superconducting strands soldered to the surface of the cable that can be installed in the ITER assembly hall and at the cold test facility. This cylinder serves as a transition area between the CS module and the bus extension.We made two racetrack samples and tested four bus joints in our Joint Test Apparatus. Resistance of the bus joints was measured by a decay method and by a microvoltmeter; the value of the current was measured by the Hall probes. This measurement method was verified in the previous tests. The resistance of the joints varied insignificantly from 1.5 to 2 nΩ.One of the challenges associated with a soldered joint is the inability to use corrosive chemicals that are difficult to clean. This paper describes our development work on cable preparation, chrome removal, compaction, soldering, and final assembly and presents the test results.     

Development of the plasma facing components in Japan for ITER

After the successful completion of the prequalification activity for ITER divertor procurement, Japanese Domestic Agency (JADA) and ITER Organization (IO) have entered into the procurement arrangement of divertor Outer Vertical Target (OVT) in June 2009. In accordance with the arrangement, JADA is going to fully procure the outer target components which correspond to 60 divertor cassettes. JADA has started to manufacture an OVT full-scale prototype in order to pick out/solve technical and quality issues and to establish a rational manufacturing process toward the start of the series of production of the OVT components to be installed in ITER. This paper presents the overview of JADA's manufacturing activity and the procurement schedule on the divertor outer target procurement.     

Development of Prototype Neutron Flux Monitor for ITER

The prototype neutron flux monitor consists of a high purity 235U fission chamber detector and a “blank” detector, which is a fissile material free detector with the same dimension as the fission chamber detector to identify noise issues such as noise coming from gamma rays. The main parameters of the fission chamber assembly that have been measured in the laboratory are confirmed to approach the technological level of the International Thermonuclear Experimental Reactor (ITER) in the near future. This prototype neutron flux monitor will be further developed to become a neutron flux monitor suitable for the operation phase of D-D fusion on the ITER.     

Verification test results of a cutting technique for the ITER blanket cooling pipes

For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.     

Development of a RF-driven ion source for the ITER NBI system

Extensive R&D work on RF-driven negative hydrogen ion sources carried out at IPP Garching led to the decision of ITER to select this type of source as the new reference source for the ITER NBI system. The principle suitability of the RF source has been demonstrated in a small scale, short pulse length experiment: accelerated current densities, co-extracted electron currents at a source operation pressure, all well inside the range of the ITER requirements have been achieved simultaneously. In subsequent experiments, pulse lengths up to 1 h and the possibility of modularly extending the source to ITER source dimensions were demonstrated. The results achieved at the various IPP test beds, the lessons learnt during optimising the source for negative ion production and extraction as well as the problems still to be solved are summarized. As the next step in support of the NBI development for ITER, IPP plans to build a new test facility for beam extraction from a source of half the size for ITER.     

Predictions for EAST Divertor Performance

A detailed study of the divertor performance in EAST has been performed for both its double null （DN） and single null （SN） configurations. The results of application of the SOLPS （B2-Eirene） code package to the analysis of the EAST divertor are summarized. In this work, we concentrate on the effects of increased geometrical closure and of magnetic topology variation on the scrape-off layer （SOL） and divertor plasma behavior. The results of numerical predictions for the EAST divertor operational window are also described in this paper. A simple Core-SOL- Divertor （C-S-D） model was applied to investigate the possibility of extending plasma operational space of low hybrid current drive （LHCD） experiments for EAST.     

Development of the supervisory systems for the ITER diagnostic systems in JADA

In ITER, it is important how the CODAC system conducts many plant systems including diagnostic systems. In order to establish necessary communications between the diagnostics systems and the CODAC system, Japan domestic agency (JADA) has proposed the new concept of supervisory system for the diagnostic system based on our experiences in operating plasma diagnostic systems. The supervisory system manages operation sequences, current state and configuration parameters for the measurement. JADA designed the supervisory system satisfying the requirements from both CODAC system and diagnostic systems. In our design, the tool which converts operational steps described as flowcharts into the EPICS (experimental physics and industrial control system) records source codes is introduced. This tool will ensure reduction of the system designers’ efforts. We designed a communication protocol to configure measurement parameters and proposed configuration parameter validation function. We also analyzed the management of the central/local control mode for the diagnostic systems. The function which selects the adequate limit values and consistency check algorithms in accordance with the conditions of the diagnostics system is proposed. JADA will develop a prototype of the supervisory system and validate the design in 2013.     

Development of ITER tritium transport package

For tritium supply to the fusion reactor of ITER (International Thermonuclear Experimental Reactor; the way to new energy) [1], tritium needs to be transported from tritium production sites, mainly the CANDU type reactor sites to the Tritium Plant building of ITER. Korea Atomic Energy Research Institute (KAERI) was commissioned the work of developing the transport package for tritium by ITER organization and the first stage of the development has been finished. The developed package was designed to carry 70 g of tritium and classified as a type B(U) package, which should comply with the requirements stipulated in IAEA regulations for the transportation of radioactive materials [2]. The package is composed of a storage vessel, a containment vessel, an overpack and an aluminum liner, which is a unique feature of the package. The aluminum liner between the storage vessel and the containment vessel is for containment control under the repetitive use of the package. The package has enough pressure resistance for 5 years in-site storage and the structural and thermal integrity under the hypothetical accident conditions has been demonstrated through a series of analyses.     

Development of the plasma facing components for the dome-liner component of the ITER divertor

On the basis of the design and the specification of the dome-liner elaborated by EFDA, a manufacturing route based on high temperature brazing has been developed and proved by means of the fabrication and testing of several samples and mock-ups. The dome is protected with tungsten armour tiles joined onto heat sinks obtained from a bimetallic plate made of precipitation hardened copper–chromium–zirconium alloy and stainless steel realized by explosion bonding. The brazed joint between the tungsten tiles and the heat sink has been qualified by means of thermal fatigue tests on small-scale mock-ups in reactor relevant conditions. The properties of the explosion bonding joint between the front copper alloy plate to the rear steel backing has been assessed by means of an extensive metallurgical and mechanical test program according to the specification provided by EFDA. The dimensional stability during the fabrication route has been investigated by means of the realization of a relevant curved component that has been dimensionally tested after the completion of each step of the manufacturing route. The results of the experimental activity are presented and discussed in this paper.     

Development of in situ cleaning techniques for diagnostic mirrors in ITER

Mirrors will be used in all optical and laser-based diagnostic systems of ITER. In the severe environment, the optical characteristics of mirrors will be degraded, hampering the entire performance of the respective diagnostics. A minute impurity deposition of 20 nm of carbon on the mirror is sufficient to decrease the mirror reflectivity by tens of percent outlining the necessity of the mirror cleaning in ITER. The results of R&D on plasma cleaning of molybdenum diagnostic mirrors are reported. The mirrors contaminated with amorphous carbon films in the laboratory conditions and in the tokamaks were cleaned in steady-state hydrogenic plasmas. The maximum cleaning efficiency of 4.2 nm/min was reached for the laboratory and soft tokamak hydrocarbon films, whereas for the hard tokamak films the carbidization of mirrors drastically decreased the cleaning efficiency down to 0.016 nm/min. This implies the necessity of sputtering cleaning of contaminated mirrors as the only reliable tool to remove the deposits by plasma cleaning. An overview of R&D program on mirror cleaning is provided along with plans for further studies and the recommendations for ITER mirror-based diagnostics.     

Development of Data Acquisition Management System for ITER DC Testing Platform

To acquire multi-channel signals with 10 kHz sampling rate from various front-end sensors, a Data Acquisition Management System (DAMS) based on MDSplus was designed for the International Thermonuclear Experimental Reactor (ITER) Direct Current (DC) testing platform. Due to a large number of experimental data generated from long-pulse operation, it is very important to view and analyze experimental data online during operation. To meet the requirement of online data processing, slice storage and thumbnail technology were applied in DAMS. The long pulse data is gradually written in MDSplus database. The DAMS has been verified in the ITER DC power supply testing platform.     

Analyses of the ITER vacuum vessel with the use of a new modelling technique

The on-going design development of the ITER vacuum vessel (VV) has been supported and accompanied recently by extensive analytical studies of its different structural aspects. The use of a newly developed parametric model of the ITER VV sector with port structures has enabled various detailed assessments of this structure. Implementation of new features in the model and new approach for its generation has eased updates and modifications of the model, reducing the time necessary to introduce them into the model and allowing it to keep pace with discussions on planned design changes.     

Dust Divertor for a Tokamak Fusion Reactor

Micron-size tungsten particulates find their equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic ion flow drag parallel to the divertor surface. The natural circulation of the dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.