Engineering evolution of the FAST machine

FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploit innovative plasma facing component systems for DEMO. FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity δ = 0.4) and cost effective machine able to investigate, with integration capability, non linear dynamics effects of alpha particle behaviour in burning plasmas. FAST operates in high performance H-mode (B_T up to 8.5 T; I_P up to 8 MA), as well as in advanced tokamak mode (I_P = 3 MA), and in full non inductive current mode (I_P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets. This allows for a pulse duration up to 170 s at 3 MA/3.5 T. The vacuum vessel (VV), segmented into 20-degree modules, is capable to accommodate a 40 MW RF power system. The machine has been designed to house a 10 MW Negative Neutral Beam Injection (NNBI) system. Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plate materials, and argon or neon as the impurities to be injected for mitigating the thermal loads.     

High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current I_P = 10 MA, toroidal field B_T = 8.5 T, with a q₉₅ ≈ 2.3 that would correspond to I_P ≈ 20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated.     

Plasma scenarios, equilibrium configurations and control in the design of FAST

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13–18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems.FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. All plasma equilibria satisfy the following constraints: (a) minimum distance of 3 energy e-folding length (assumed to be 1 cm on the equatorial plane) between plasma and first wall to avoid interaction between plasma and main chamber; (b) maximum current density in the poloidal field coils, transiently, up to around 30 MA/m². The discharge duration is always limited by the heating of the toroidal field coils that are inertially cooled by helium gas at 30 K. The location of the poloidal field coils has been optimized in order to: minimize the magnetic energy; produce enough magnetic flux (up to 35 Wb stored) for the formation and sustainment of each scenario; produce a good field null at the plasma break-down (B_P/B_T < 2 × 10⁻⁴ at low field, i.e. B_T = 4 T and E_T = 2 V/m for at least 40 ms).Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100 ms.     

Measurement of L X-ray fluorescence cross-sections for elements with 45 ? Z ? 50 using synchrotron radiation at 8 keV

The L shell fluorescence cross-sections of the elements in range 45 ? Z ? 50 have been determined at 8 keV using Synchrotron radiation. The individual L X-ray photons, Ll, Lα, Lβ_I, Lβ_II, Lγ_I and Lγ_II produced in the target were measured with high resolution Si(Li) detector. The experimental set-up provided a low background by using linearly polarized monoenergetic photon beam, improving the signal-to-noise ratio. The experimental cross-sections obtained in this work were compared with available experimental data from Scofield [1] and [2] Krause [3] and [4] and Scofield and Puri et al. [5] and [6].These experimental values closely agree with the theoretical values calculated using Scofield and Krause data, except for the case of Lγ, where values measured of this work are slighter higher.     

A Fusion Nuclear Science Facility for a fast-track path to DEMO

An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 m, κ = 2.3, B_T = 5.4 T, I_P = 6.6 MA, β_N = 2.75, P_fus = 127 MW. The modest bootstrap fraction of ƒ_BS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ∼ 10 in ITER.     

The three RF additional heating systems for FAST

FAST (Fusion Advanced Studies Torus) is a new tokamak machine proposed by the Italian Fusion Association as a Satellite Tokamak for the ITER programme. FAST will operate with deuterium plasmas to avoid the complexity deriving from the use of tritium. Therefore burning plasma conditions, where energy density of fast ions and of charged fusion products is a significant fraction of the total plasma energy density, will be achieved by accelerating plasma ions above the half-MeV range through an Ion Cyclotron Resonance Heating (ICRH) system (P = 30 MW, f = 60–90 MHz). For long pulse Advanced Tokamak (AT) scenarios, a Lower Hybrid Current Drive (LHCD) system (P = 6 MW, f = 3.7 GHz) has been envisaged to actively control the current profile, whereas an Electron Cyclotron Resonant Heating (ECRH) system (P = 4 MW, f = 170 GHz) will provide enough RF power for MHD control.     

Conceptual design study of fusion DEMO plant at SWIP

The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m² is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li₄SiO₄), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced.     

FAST: A European ITER satellite experiment in the view of DEMO

Fusion Advanced Studies Torus (FAST) aims to contribute to the exploitation of ITER and to explore innovative DEMO technology. FAST has been designed to study, in an integrated scenario: (a) relevant plasma-wall interaction problems, with a large power load (P/R ～ 22 MW/m; P/R² ～ 12 MW/m²) and with a full metallic wall; (b) to tackle operational problems in regimes with relevant fusion parameters; (c) to investigate the non-linear dynamics of fast particles (alpha like) in burning plasmas. FAST will operate on a wide parameters range, namely in high performance H-mode (BT ～ 8.5 T; IP ～ 8 MA) as well as in advanced Tokamak operation up to full non-inductive current scenario (IP ～ 2 MA). The main heating is based on 30 MW ICRH, but the ports have been designed to allocate up to 20 MW of 1 MeV NNBI. Helium gas at 30 K is used for cooling of the full machine, a preliminary analysis shows the possibility of realizing FAST with a complete superconductor set of coils. An innovative active system is under development to reduce and to control the magnetic ripple. Tungsten (W) or liquid lithium (L–Li) has been chosen for the divertor material plates and the code EDGE2D has been used to optimize the divertor geometry.     

Overview of engineering design,manufacturing and assembly of JT-60SA machine

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R&D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.     

Design and experimental results of feedback control of Ohmic-heating transformer magnetic flux by LHCD power in HT-7 Tokamak

In order to make a research on long pulse or even steady state operation with non-inductive drive in plasma discharge, a new feedback control scheme instead of the previous one has been designed and operated in HT-7 [HT-7 team presented by J. Li, et al., Plasma Phys. Control. Fusion 42 (2) (2000) 135-146] Tokamak experiment, 2004. Consumption of iron-core transformer magnetic flux (MFT) is feedback controlled for the first time by power of lower hybrid current drive (LHCD) P_LH, when the Ohmic-heating circuit current can maintain the plasma current I_P constant with another feedback control loop, which make MFT evolve at alternating-change state to avoid flux saturation. Plasma current I_P can be maintained steadily up to 120 s in this operation mode at reduced plasma parameters (I_P ≈ 50-100 KA, average density , P_LH = 100-200 KW). Design and experimental results are presented in the paper, which including control model analysis, configurations of control system and MFT feedback control experiments in HT-7. The high voltage power supply (HVPS) of LHCD is the main controller that regulates the LHCD power into the plasma to control the MFT.     

Nuclear analysis of DEMO water-cooled blanket based on sub-critical water condition

For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with P_n is 5 MW/m². Furthermore, the local TBR was optimized by changing the material proportion for each P_n level (1-5 MW/m²). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with P_n level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the P_n level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.     

The first performance of the facility for testing ITER CC conductor short sample

A new facility had been set up to test the low temperature properties of the short sample of the small-size cable-in-conduit conductor (CICC). The facility consisted of the background magnet which could provide 7 T centric magnetic field, a 50 kA superconducting transformer which provided sample current, a 500 W/4.5 K helium refrigerator which produced both the liquid helium (LHe) and supercritical helium (SHe). An ITER CC conductor short sample was prepared and measured in this testing system. T_cs of 7.02 K (@4.1 T, 10 kA) and I_c of 8.9 kA (@4.1 T, 7.06 K) were measured.     

ITER operational space for full plasma current H-mode operation

Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I_p^{^}4. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I_p of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.     

The first DC performance test and analysis of CC conductor short sample at ASIPP conductor test facility

The first DC performance experiments of ITER correction coil (CC) conductor short sample have been carried out in the conductor test facility of Institute of Plasma Physics, CAS (ASIPP) in January this year. Those experiments aim to investigate the DC performance of ITER CC conductor. The tested conductor short sample is bended as a half circle with the diameter of 270 mm to meet the background magnetic field shape. The half circle part of sample is longer than the final twist pitch. The current sharing temperature (T_cs) in the 3.86 T external magnetic field (B_ex), ≤12 kA could be measured including the critical current (I_c) run. There is no obvious impact of 1000 cycles on DC performance. Those measured T_cs results are in agreement with the expected results from strand scaling.     

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.     

Simulations of the SOL plasma for FAST,a proposed ITER satellite tokamak

The first simulations with EDGE2D/EIRENE code of the SOL plasma in the FAST tokamak have been run for the basic H-mode scenario. Its similarity to ITER and relevance for DEMO bring interest to the study. Five different preliminary divertor designs have been examined by varying density at separatrix over the plausible range n_s,out = 0.7–1.0 × 10²⁰ m^?3. Margins exist for optimizing the design and minimizing the impurity injection rate even at the lowest density, with load below the safe limit of 18 MW/m² on the monoblock W targets, and to achieve a good degree of detachment at higher density. Both the plate tilting angle and the neutral dynamics are crucial factors. The detachment level can be significantly increased for the higher density scenario, while for the full non-inductive operation the injection of impurities will probably be necessary to reduce the heat load.     

Measurement of thermal neutron cross-sections and resonance integrals for Hf(n,γ)Hf and Hf(n,γ)Hf reactions at the Pohang neutron facility

The thermal-neutron cross-sections and the resonance integrals for the ¹⁷⁹Hf(n,γ)^180mHf and the ¹⁸⁰Hf(n,γ)¹⁸¹Hf reactions have been measured by the activation method. The high purity Hf and Au metallic foils within and without a Cd shield case were irradiated in a neutron field of the Pohang neutron facility. The gamma-ray spectra from the activated foils were measured with a calibrated p-type high-purity Ge detector.In the experimental procedure, the thermal neutron cross-sections, σ₀, and resonance integrals, I₀, for the ¹⁷⁹Hf(n,γ)^180mHf and the ¹⁸⁰Hf(n,γ)¹⁸¹Hf reactions have been determined relative to the reference values of the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction, with σ₀ = 98.65 ± 0.09 barn and I₀ = 1550 ± 28 barn. In order to improve the accuracy of the experimental results, the interfering reactions and necessary correction factors were taken into account in the determinations. The obtained thermal neutron cross-sections and resonance integrals were σ₀ = 0.424 ± 0.018 barn and I₀ = 6.35 ± 0.45 barn for the ¹⁷⁹Hf(n,γ)^180mHf reaction, and σ₀ = 12.87 ± 0.52 barn and I₀ = 32.91 ± 2.38 barn for the ¹⁸⁰Hf(n,γ)¹⁸¹Hf reaction. The present results are in good agreement with recent measurements.     

Recent progress on the Tokamak Fusion Test Reactor

The deuterium-tritium (D-T) experiments on the Tokamak Fusion Test Reactor (TFTR) have yielded unique information on the confinement, heating and alpha particle physics of reactor scale D-T plasmas as well as the first experience with tritium handling and D-T neutron activation in an experimental environment. The D-T plasmas produced and studied in TFTR have peak fusion power of 10.7 MW with central fusion power densities of 2.8 MWm^–3 which is similar to the 1.7 MWm^–3 fusion power densities projected for 1,500 MW operation of the International Thermonuclear Experimental Reactor (ITER). Detailed alpha particle measurements have confirmed alpha confinement and heating of the D-T plasma by alpha particles as expected. Reversed shear, highl _i and internal barrier advanced tokamak operating modes have been produced in TFTR which have the potential to double the fusion power to 20 MW which would also allow the study of alpha particle effects under conditions very similar to those projected for ITER. TFTR is also investigating two new innovations, alpha channeling and controlled transport barriers, which have the potential to significantly improve the standard advanced tokamak.     

A remotely steered millimetre wave launcher for electron cyclotron heating and current drive on ITER

High-power millimetre wave beams employed on ITER for heating and current drive at the 170 GHz electron cyclotron resonance frequency require agile steering and tight focusing of the beams to suppress neoclassical tearing modes. This paper presents experimental validation of the remote steering (RS) concept of the ITER upper port millimetre wave beam launcher. Remote steering at the entrance of the upper port launcher rather than at the plasma side offers advantages in reliability and maintenance of the mechanically vulnerable steering system. A one-to-one scale mock-up consisting of a transmission line, mitre bends, remote steering unit, vacuum window, square corrugated waveguide and front mirror simulates the ITER launcher design configuration. Validation is based on low-power heterodyne measurements of the complex amplitude and phase distribution of the steered Gaussian beam. High-power (400 kW) short pulse (10 ms) operation under vacuum, diagnosed by calorimetry and thermography of the near- and far-field beam patterns, confirms high-power operation, but shows increased power loss attributed to deteriorating input beam quality compared with low-power operation. Polarization measurements show little variation with steering, which is important for effective current drive requiring elliptical polarization for O-mode excitation. Results show that a RS range of up to −12° to +12° can be achieved with acceptable beam quality. These measurements confirm the back-up design of the ITER ECRH&CD launcher with future application for DEMO.