Status of the IDTF high-heat-flux test facility

The ITER Divertor Test Facility (IDTF) was designed for the high heat flux tests of outer vertical targets, inner vertical targets and domes of the ITER divertor. This facility was created in the Efremov Institute under the Procurement Arrangement 1.7.P2D.RF (high heat flux tests of the plasma facing units of the ITER divertor).The heat flux is generated by an electron-beam system (EBS), 800 kW power and 60 kV maximum accelerating voltage. The component to be tested is mounted on a manipulator in the vacuum chamber capable of testing objects up to 2.5 m long and 1.5 m wide. The pressure in the vacuum chamber is about 3*10⁻³ Pa. The parameters of the cooling system and the water quality (deionized water) are similar to the cooling conditions of the ITER divertor. The integrated control system regulates all IDTF subsystems and data acquisition from all diagnostic devices, such as pyrometers, IR-cameras, video cameras, flow, pressure and temperature sensors.Started in 2008, the IDTF was commissioned in 2012 with the testing the outer vertical full-scale prototypes and the completion of the PA 1.7.P2D.RF task. This paper details the main characteristics of the IDTF.     

ASDEX Upgrade enhancements in view of ITER application

The aim of the ASDEX Upgrade (AUG) programme is to support the design, prepare the physics base and develop regimes beyond the baseline of ITER and for DEMO. Its ITER-like geometry, poloidal field system, versatile heating system and power fluxes make AUG particularly suited.After the transition to fully tungsten coated plasma facing components AUG could be operated without prior boronizations and a low permanent deuterium retention was found qualifying W as wall material. ITER-like baseline H-modes (H₉₈ ～ 1, β_N ～ 2) were routinely achieved up to 1.2 MA plasma currents. W concentrations could be kept at an acceptable level of <5 × 10^?5 by central wave heating (enhancing impurity outward transport) and ELM pacing with gas puffing. The compatibility of high performance improved H-modes, the ITER hybrid scenario, with an un-boronized W wall was demonstrated achieving H₉₈ ～ 1.1 and β_N up to 2.6 at modest triangularities δ ≤ 0.3. This performance is reached despite the gas puffing needed for W influx control. Increasing δ to 0.35 allowed at even higher puff rates still a H₉₈ ～ 1.1.Reliable plasma operation in support of ITER comprised the demonstration of ECRF assisted low voltage plasma start-up and current rise at toroidal electric fields below 0.3 V/m resulting in a ITER compatible range of plasma internal inductance of 0.71–0.97. Disruption mitigation is feasible using strong gas puffs, and the achieved electron densities approach values needed for runaway suppression.Present hardware extensions in support of ITER include the upgrading of ECRH by a 4 MW/10 s system with large deposition variability (tuneable frequency between 105 and 140 GHz, real-time steerable mirrors) for central heating and MHD mode control. A powerful system of 24 in-vessel coils produces error fields up to toroidal mode number n = 4 for ELM suppression and mode rotation control. In connection with a close conducting wall they will open up the road for RWM stabilization in advanced scenarios. For those we are considering LHCD for current drive and profile control with up to 500 kA driven current. The tungsten sources are dominated by sputtering from intrinsic light impurities, and the W influx from the outboard limiters are the main source for the core plasma. ICRH induced electric fields accelerate light impurities, restricting the use of ICRH to just after boronization. 4-strap antennas imbedded in extended wall structures might solve this problem. Finally, doubling the plasma volume with plasma currents above 2 MA in AUG could be the solution for a needed ITER satellite.     

Parametric study of one triplet of the ITER ICRH antenna by numerical modeling

Each of the two ITER ICRF antennas consists of a close-packed array of 24 straps arranged in a 6 poloidal by 4 toroidal array. Three poloidally adjacent straps (a “triplet” of straps) are fed together through a 4-port junction from one 20 Ohm feeding line. The complete array has to radiate 20 MW of RF power over a frequency range of 40 MHz to 55 MHz and for different toroidal phasings. The RF optimization of the antenna has been performed numerically on one triplet of straps (1/8th of the antenna) [1], [2]. In parallel a number of reduced-scale mock-ups of one triplet of the ITER ICRH antenna were constructed in order to validate the results of the numerical optimization [1], [3].The aim of this work is primarily to benchmark the CST MWS^® [4] numerical modeling against numerous measurements done on the mock-up of the 2007 design. Moreover MWS calculates the 3D distribution of the currents and of the fields of the triplet. Hence it gives the possibility to check the fields and current distributions resulting from the optimisation study of the ITER ICRH antenna triplet done by changing geometrical parameters of the straps and antenna box of the mock-up of 2007 design [1], [2], [3]. The considered parameters are: strap width, antenna box depth and vertical septum recess with respect to the front of the current strap. The impact of the presence of the Faraday screen is also evaluated.Excellent agreement between modeled and measured S parameters is obtained. Analysis of the fields and currents distributions on the straps is reported. Excellent current balance is confirmed.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Operational experience with forced cooled superconducting magnets

Force-cooled concept has been chosen for ITER superconducting magnet to get reliable coil insulation using vacuum-pressure impregnation (VPI) technology. However 17 breakdowns occurred during operation of six magnets of this type or their single coil tests at operating voltage < 3 kV, while ITER needs 12 kV. All the breakdowns started on electric, cryogenic and diagnostic communications (ECDCs) by the high voltage induced at fast current variations in magnets concurrently with vacuum deterioration, but never on the coils, though sometimes the latter were damaged too. It suggests that simple wrap insulation currently employed on ECDCs and planned to be used in ITER is unacceptable. Upgrade of the ECDC insulation to the same level as on the coils is evidently needed. This could be done by covering each one from ECDCs with vacuum-tight grounded stainless steel casings filled up with solid insulator using VPI-technology. Such an insulation will be insensitive to in-cryostat conditions, excluding helium leaks and considerably simplifying the tests thus allowing saving time and cost. However it is not accepted in ITER design yet. So guarantee of breakdown prevention is not available.     

Overview of results from Wendelstein7-X HTS current lead testing

The High Temperature Superconductor (HTS) current leads (CL) for the Wendelstein 7-X stellarator (W7-X) with a maximum current of 18.2 kA are designed and manufactured by the Karlsruhe Institute of Technology (KIT). In addition the acceptance tests of the W7-X HTS CLs are performed at KIT. Therefore the existing TOSKA facility has been extended by a test cryostat connected to the main vacuum vessel. After the extensive prototype CL test campaigns in 2010 the final acceptance tests of 14 series CLs started in 2011. The estimated completion of the routine test campaign is in December 2012. The main parts of each acceptance test are the determination of the heat load at the 4.5 K level, of the necessary 50 K He mass flow rate through the heat exchanger as well as the simulation of a loss of flow accident of the 50 K He mass flow at full current (18.2 kA). The tests also include a long-time operation at the maximum current of 18.2 kA to demonstrate the steady state operation capability of the HTS CLs. In the present paper an overview of all conducted HTS CL acceptance tests is given. The results for the different CLs are summarized and compared to the specifications.     

Feeder components and instrumentation for the JT-60SA magnet system

The modifying of the JT-60U magnet system to the superconducting coils is progressing as a satellite facility for ITER by both parties of Japanese government and European commission in the Broader Approach agreement. The magnet system requires current supplies of 25.7 kA for 18 TF coils and of 20 kA for 4 CS modules and 6 EF coils. The magnet system generates an average heat load of 3.2 kW at 4 K to the cryogenic system. The feeder components connected to the power supply provide current supply. The cooling pipes connected to the cryogenic system provide coolant supply. The instrumentation of the JT-60SA magnet system is used for its operation.     

High temperature superconductors for the ITER magnet system and beyond

The use of high temperature superconductor (HTS) materials in future fusion machines could increase the efficiency drastically, but strong boundary conditions exist. To outline the prospects, challenges and problems, first the benefit of using HTS materials is estimated considering the saving in cryogenic power. Next, it is demonstrated that industrial available HTS materials can be used for fusion today. For this purpose, we give a short summary of results that have been obtained from an ITER conform 70 kA HTS current lead that was designed, built and tested by the Forschungszentrum Karlsruhe and the CRPP Villigen in the frame of the European Fusion Technology Programme and in cooperation with industry. This current lead consists of an HTS part that covered the temperature range from 4.5 to 70 K and a conventional part, making the connection to room temperature. Because the HTS part had no ohmic losses and poor thermal conduction, the refrigerator power necessary for cooling the current lead was reduced drastically. The saving factor could be calculated to be 5.4 at zero current and 3.7 at 68 kA. The current lead could even be operated at 80 kA and with respect to safety criteria of ITER, a complete loss of He flow was simulated showing that the HTS current lead could hold a current of 68 kA for 6 min without active cooling. These results demonstrate that today existing HTS materials can be used in ITER for current leads or bus bar systems.For fusion machines beyond ITER, the development of an HTS fusion conductor would be the key to operate the complete magnet system at higher temperatures. The option of developing fusion conductors based on Bi-2223 and YBCO are briefly discussed. For a success of such conductors, the AC loss optimisation is crucial.     

Prospects for the use of high-Tc superconductors in fusion magnets and options for their test in SULTAN

In the last few years, the critical current densities of long commercially available REBa₂Cu₃O_7?x (RE-123, where RE represents Y or a rare earth element) coated conductors have reached values of 250 A/cm-width at 77 K and zero applied field. Even higher values of 600 A/cm-w (77 K, B = 0) have been demonstrated in shorter lengths. The attractive features of the use of these high-T_c superconductors (HTS) are operation temperatures above 20 K and/or magnetic fields higher than those envisaged for the ITER TF coils. Possible operation conditions for HTS fusion magnets have been studied taking into consideration the possible further improvements of RE-123 coated conductors. Investigations of stability and quench behavior indicate that stability is not a problem, whereas quench detection and protection need attention. Because of the high currents necessary for fusion magnets, many tapes need to be assembled into a transposed conductor. The qualification of HTS conductors for fusion magnets would require their test at magnetic fields of 11 T and currents well above 10 kA. The possibilities to test straight HTS conductor samples in SULTAN have been considered. For a test at 4.5 K, only the development of a low resistance joint between the HTS conductor under test and the NbTi transformer of SULTAN would be necessary. Tests up to 20 K would require that the HTS sample is connected with the NbTi transformer by a conduction-cooled HTS bus bar of large thermal resistance similar to the HTS module of a current lead. HTS conductor tests at temperatures around 50 K would be possible with modified cryogenics.     

Manufacturing,testing and post-test examination of ITER divertor vertical target W small scale mock-ups

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets.In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes.According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m² and finally 1000 cycles at 20 MW/m².After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing.A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area.The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks.This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.     

Power dissipation and energy transfer during testing of the ITER toroidal field model coil

The testing of the ITER toroidal field model coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the LCT coil and 80 kA in the TFMC, respectively. The heat load of both coils, including the eddy current losses in the passive structures and the joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. A computer code, based on the full inductance and resistance matrices, has been developed with SIMULINK™. After validation with experimental data the code has been used to perform circuit analysis and to evaluate the power dissipation and energy transferred to the cryogenic plant and to the external power circuits.     

Qualification tests for ITER TF conductors in SULTAN

From February 2007 to May 2008, 18 short length conductor sections have been tested in SULTAN for design verification and manufacturer qualification of the ITER Toroidal Field (TF) conductor. The test program is focussed on the current sharing temperature, T_cs, at the nominal operating conditions, 68 kA current and 11.15 T effective field, which can be fully reproduced in the SULTAN test facility. A broad range of results was observed, with over 2 K difference among the T_cs of the conductors. In average, the results are poorer compared to the potential performance estimated from the strand scaling law. The key parameters to mitigate the degradation are not yet clearly identified. The experimental challenges to test conductors with performance degradation are highlighted, including enhanced instrumentation sets, the application of gas flow calorimetry to sense the current sharing power and the post-processing of voltage data to cancel the transverse potential across the cable. The updated schedule of the tests in SULTAN is presented with the short-term action plan for conductor test.     

Study of large gap 200 kV breakdowns in vacuum with high stored energy for ITER neutral beam accelerator

In the context of the ITER contract “ITER/CT/07/219–200 kV Stored Energy Tests”, electrical breakdown tests have been performed in vacuum with a stored energy of up to 425 J. The experiments have been conceived and performed with the collaboration of Consorzio RFX. The tests are being performed in the 1 MV test facility at IRFM, CEA-Cadarache. They should simulate the conditions that will be found in the ITER Neutral Beam accelerator, at 200 kV. This paper presents the set-up of the test bed, the choice of critical components, the diagnostic equipments and the results obtained with 200 kV applied on the anode electrode.     

Updating the design of the feeder components for the ITER magnet system

The ITER superconducting magnet system generates an average heat load of 23 kW at 4 K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68 kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0 MPa and temperature between 4.3 and 4.7 K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components.     

Recent improvements to the ITER neutral beam system design

The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium–tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D⁰ or up to 870 keV H⁰) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3], [4], [5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6], [7], [8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community.     

The performance test and analysis of ITER Main and Correction Busbar conductor

The first ITER Main Busbar (MBCN1) and Correction Busbar (CBCN1) conductor samples were manufactured in ASIPP and tested in the SULTAN facility. This paper introduces the sample manufacture, including strand, cabling, jacketing and sample preparation, and discusses the performance of MBCN1 and CBCN1 conductors. The testing results show that both samples have high T_cs, and meet the ITER requirement.Due to the ITER acceptance standard T_cs of MB conductor was changed to 6.7 K at 45.5 kA/3.9 T. The performance of MBCN1 conductor after cyclic load fits the ITER requirement, but the sample was only tested at 57 kA/2.75 T before cycling test. Using some hypothesis and equation to extrapolate the T_cs performance of MBCN1 conductor before cycling test, the result also fits the ITER requirement.For CBCN1 conductor, the central line of the central cooling spiral shifted about 1.3 mm during the cabling. The deviation causes an increase of the max self-field by about 0.005 T, which could not influence the CBCN1 conductor real T_cs performance at peak field.     

Proposed high voltage power supply for the ITER relevant lower hybrid current drive system

In the framework of the EFDA task HCD-08-03-01, the ITER lower hybrid current drive (LHCD) system design has been reviewed. The system aims to generate 24 MW of RF power at 5 GHz, of which 20 MW would be coupled to the plasmas. The present state of the art does not allow envisaging a unitary output of the klystrons exceeding 500 kW, so the project is based on 48 klystron units, leaving some margin when the transmission lines losses are taken into account. A high voltage power supply (HVPS), required to operate the klystrons, is proposed. A single HVPS would be used to feed and operate four klystrons in parallel configuration. Based on the above considerations, it is proposed to design and develop twelve HVPS, based on pulse step modulator (PSM) technology, each rated for 90 kV/90 A. This paper describes in details, the typical electrical requirements and the conceptual design of the proposed HVPS for the ITER LHCD system.     

The ITER full size plasma source device design

In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m² in D₂ (D^?) and 350 A/m² in H₂ (H^?) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m².The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.     

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.     

Commissioning and initial operation of KSTAR superconducting tokamak

The commissioning and the initial operation for the first plasma in the KSTAR device have been accomplished successfully without any severe failure preventing the device operation and plasma experiments. The commissioning is classified into four steps: vacuum commissioning, cryogenic cool-down commissioning, magnet system commissioning, and plasma discharge.Vacuum commissioning commenced after completion of the tokamak and basic ancillary systems construction. Base pressure of the vacuum vessel was about 3 × 10^?6 Pa and that of the cryostat about 2.7 × 10^?4 Pa, and both levels meet the KSTAR requirements to start the cool-down operation. All the SC magnets were cooled down by a 9 kW rated cryogenic helium facility and reached the base temperature of 4.5 K in a month. The performance test of the superconducting magnet showed that the joint resistances were below 3 nΩ and the resistance to ground after cool-down was over 1 GΩ. An ac loss test of each PF coil made by applying a dc biased sinusoidal current showed that the coupling loss was within the KSTAR requirement with the coupling loss time constant less than 35 ms for both Nb₃Sn and NbTi magnets. All the superconducting magnets operated in stable without quench for long-time dc operation and with synchronized pulse operation by the plasma control system (PCS). By using an 84 GHz ECH system, second harmonic ECH assisted plasma discharges were produced successfully with loop voltage of less than 3 V. By the real-time feedback control, operation of 100 kA plasma current with pulse length up to 865 ms was achieved, which also meet the first plasma target of 100 kA and 100 ms. The KSTAR device will be operated to meet the missions of steady-state and high-beta achievement by system upgrades and collaborative researches.