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 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.     

RF modeling of the ITER-relevant lower hybrid antenna

In the frame of the EFDA task HCD-08-03-01, a 5 GHz Lower Hybrid system which should be able to deliver 20 MW CW on ITER and sustain the expected high heat fluxes has been reviewed. The design and overall dimensions of the key RF elements of the launcher and its subsystem has been updated from the 2001 design in collaboration with ITER organization. Modeling of the LH wave propagation and absorption into the plasma shows that the optimal parallel index must be chosen between 1.9 and 2.0 for the ITER steady-state scenario. The present study has been made with n_|| = 2.0 but can be adapted for n_|| = 1.9. Individual components have been studied separately giving confidence on the global RF design of the whole antenna.     

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.     

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.     

Boron lattice site location in (BGa)As and (BGa)P thin films studied using RBS and NRA with a channeled He+ ion beam

In this study the boron lattice site location in ternary B_xGa_1?xAs and B_xGa_1?xP thin films grown on (0 0 1) GaAs and (0 0 1) GaP, respectively, using low pressure metal-organic vapour-phase epitaxy (MOVPE) with boron concentrations between x = 0.8% and x = 3.2% was investigated with RBS and the ¹⁰B(α,p)¹³C nuclear reaction using a 2.3 MeV He⁺ ion beam. For this purpose, the ion beam was aligned with the [0 0 1], [0 1 1] and [1 1 1] axis and the RBS and proton yield from the nuclear reaction compared with random ion incidence. For comparison, theoretical proton yields which assume boron to be located on substitutional lattice sites only were calculated for each sample/axis combination and compared with the experimental yields. The RBS/channeling measurements show a very good crystal quality of the films with χ_min being in the range of 3–5% for the [0 1 1] axis. The best crystal qualities, i.e. the lowest χ_min values and dechanneling rates, are achieved for low boron concentrations. From NRA/channeling it can be deduced that in the B_xGa_1?xAs films the fraction of interstitial boron is approximately 5% for low boron concentrations of x = 1% and 6–10% for concentrations up to x = 3.2%, whereas the fraction of interstitial boron is less than 3% in the B_xGa_1?xP film studied despite a concentration of x = 2.0%. This indicates that antisite effects of the boron incorporation are more likely in GaAs compared to GaP.     

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.     

Characterization and qualification of advanced insulators for fusion magnets

Intensive research over the past decades demonstrated that the mechanical material performance of epoxy based glass fiber reinforced plastics, which are normally used by industry as insulating materials in magnet technology, degrades dramatically upon irradiation to fast neutron fluences above 1 × 10²² m^?2 (E > 0.1 MeV). which have to be expected in large fusion devices like ITER. This triggered an insulation development program based on cyanate ester (CE) and blends of CE and epoxies, which are not affected up to twice this fluence level, and therefore appropriate for large fusion magnets like the ITER TF coils. Together with several suppliers resin mixtures with very low viscosity over many hours were developed, which renders them suitable for the impregnation of very large volumes. This paper reports on a qualification program carried out during the past few years to characterize suitable materials, i.e. various boron-free R-glass fiber reinforcements interleaved with polyimide foils embedded in CE/epoxy blends containing 40% of CE, a repair resin, a conductor insulation, and various polyimide/glass fiber bonded tapes. The mechanical properties were assessed at 77 K in tension and in the interlaminar shear mode under static and dynamic load conditions prior to and after reactor irradiation at ~340 K to neutron fluences of up to 2 × 10²² m^?2 (E > 0.1 MeV). i.e. twice the ITER design fluence. The results confirmed that a sustainable solution has become available for this critical magnet component of ITER.     

Molecular-dynamics simulation of Si1−xGex epitaxial growth on Si(100)

Molecular dynamics study of the Si_1−xGe_x epitaxial growth on Si(1 0 0) substrate utilizing the Stillinger–Weber two- and three-body interaction potentials was carried out. The Stranski–Krastanov growth mechanism of the Si_1−xGe_x strain layers on Si(1 0 0) was studied and compared with experiment results. The influence of different x on the epitaxial growth layers morphology was investigated. The structure properties of the Si_1−xGe_x layers were evaluated.     

Zeff first measurements in EAST with a multi-channel visible bremsstrahlung new system

A multi-channel visible bremsstrahlung measurement system was developed to measure the ion effective charge (Z_eff) in EAST tokamak. The system has a temporal resolution of 0.05 ms and spatial resolution of 3 cm. The measurement principle and the design of the 8-channel fiber-photomultiplier tubes (PMTs) coupled system are described, including the calibration process of the measurement system with an integrating sphere. Preliminary experimental results of line integrated bremsstrahlung profile and Z_eff derived from the system are reported.     

Recent contribution of JET to the ITER physics

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δB_T = 0.08% up to δB_T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β_N values above the ‘no-wall limit’ (β_N ～ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.     

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.     

A mixed P1–DP0 diffusion theory for planar geometry

A variational analysis is used to derive a mixed P₁–DP₀ (P₁ spherical harmonics–double P₀ spherical harmonics) angular approximation to the time-independent monoenergetic neutron transport equation in one-dimensional planar geometry. This mixed angular approximation contains a space-dependent weight factor α(x) that controls the local angular approximation used at a spatial point x: α(x) = 1 yields the standard P₁ (diffusion) approximation, α(x) = 0 gives the standard DP₀ approximation, and 0 < α(x) < 1 produces a mixed P₁–DP₀ angular approximation. The diffusion equation obtained differs from the standard P₁ diffusion equation only in the definition of the diffusion coefficient. Standard Marshak incident angular flux boundary conditions are also obtained via the variational analysis. We examine the use of this mixed angular approximation coupled with the standard P₁ approximation to more accurately treat material interfaces and vacuum boundaries. We propose a simple but effective functional form for the weight factor α(x) that removes the need for the user to specify the value. Numerical results from several test problems are presented to demonstrate that significant improvements in accuracy can be obtained using this method with essentially no computational penalty.     

Gamma irradiation induced damage creation on the cation distribution,structural and magnetic properties in Ni–Zn ferrite

Ferrite of system, namely Ni_1?xZn_xFe₂O₄ with x = (0.0, 0.2, 0.4, 0.6, 0.8, 1.0), have been prepared by solid state reaction to investigate the effect of gamma rays irradiations using Co⁶⁰ source on the cation distribution, structural and magnetic properties. The unirradiated and irradiated samples were then subjected to characterization techniques such as X-ray diffraction, magnetization and AC susceptibility. The results of these characterizations are found to be different for irradiated from that of the pristine sample. The modifications in respect of irradiated samples are explained in terms of the ion-induced disorder. The important result of γ-irradiation on the cation distribution, structural and magnetic properties is the change of ratio Fe²⁺/Fe³⁺. Possible reasons on the results are proposed.     

Preliminary design of the ITER ECH Upper Launcher

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs).In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design     

Experimental investigation on cryogenic hydrogen adsorption of molecular sieves

Tritium extraction system (TES) is one of the most important components in the helium cooled solid breeder test blanket modules (TBMs) of ITER. TES will extract various isotopic species of hydrogen by the liquid nitrogen cooled molecular sieve adsorber beds (MSB). The cryogenic hydrogen adsorption properties of several kinds of molecular sieves have been investigated at the pressure of hydrogen of 100 Pa, 200 Pa, and 0.2 MPa in order to offer the suitable molecular sieve for the MSB in TES. The saturated hydrogen adsorption capacities of the MS5A-2 and MS13X-2 have been measured at 100 Pa hydrogen pressure. To demonstrate the hydrogen extraction from continuous He–H₂ purge gases, the MS5A-2 has been tested in circulating 99.79% He–0.21% H₂ mixture with a flow rate of 16.8 L/min. The results show that the globular MS5A-2 with a diameter of 3–5 mm can adsorb/desorb hydrogen quickly. The saturated hydrogen adsorption capacity of MS5A-2 is 7.55 ml g^?1 (NTP) and MS5A-2 could effectively extract trace hydrogen from mixture gases. As a result, this type of molecular sieve can be the candidate of the one in the MSB in ITER TBM.     

Optical contrast formation in amorphous silicon carbide with high-energy focused ion beams

Thin films (d ～ 1 μm) of hydrogenated amorphous silicon carbide (a-Si_1?xC_x:H), deposited by RF reactive magnetron sputtering with different carbon content x, have been implanted with high fluences (Φ = 1016–1017 cm^?2) of high-energy (E = 0.2–1 MeV) He⁺ ions as the implant species. The induced structural modification of the implanted material results in a considerable change of its optical properties, best manifested by a significant shift of the optical absorption edge to lower photon energies as obtained from photo-thermal-deflection spectroscopy (PDS) data. This shift is accompanied by a remarkable increase of the absorption coefficient over one order of magnitude (photo-darkening effect) in the measured photon energy range (0.6–3.8 eV), depending on the ion fluence, energy and carbon content of the films. These effects could be attributed both to additional defect introduction and increased graphitization, as confirmed by Raman spectroscopy and infra-red (IR) optical transmission measurements. The optical contrast thus obtained (between implanted and unimplanted film material) could be made use of in the area of high-density optical data storage using focused high-energy He⁺ ion beams.     

MeV Si ions bombardment effects on the thermoelectric properties of Co0.1SbxGey thin films

We have grown three different monolayer Co_0.1Sb_xGe_y (x = 2, 4, 11 and y = 15, 7, 15) thin films on silica substrates with varying thickness between 100 and 200 nm using electron beam deposition. The high-energy (in the order of 5 MeV) Si ion bombardments have been performed on samples with varying fluencies of 1 × 10¹², 1 × 10¹³, 1 × 10¹⁴ and 1 × 10¹⁵ ions/cm². The thermopower, electrical and thermal conductivity measurements were carried out before and after the bombardment on samples to calculate the figure of merit, ZT. The Si ions bombardment caused changes on the thermoelectric properties of films. The fluence and temperature dependence of cross plane thermoelectric parameters were also reported. Rutherford backscattering spectrometry (RBS) was used to analyze the elemental composition of the deposited materials and to determine the layer thickness of each film.