P. Andrew

High fusion performance from deuterium-tritium plasmas in JET

High fusion power experiments using DT mixtures in ELM-free H mode and optimized shear regimes in JET are reported. A fusion power of 16.1 MW has been produced in an ELM-free H mode at 4.2 MA/3.6 T. The transient value of the fusion amplification factor was 0.95±0.17, consistent with the high value of nDT(0)τEdiaTi(0) = 8.7 × 1020±20% m-3 s keV, and was maintained for about half an energy confinement time until excessive edge pressure gradients resulted in discharge termination by MHD instabilities. The ratio of DD to DT fusion powers (from separate but otherwise similar discharges) showed the expected factor of 210, validating DD projections of DT performance for similar pressure profiles and good plasma mixture control, which was achieved by loading the vessel walls with the appropriate DT mix. Magnetic fluctuation spectra showed no evidence of Alfvenic instabilities driven by alpha particles, in agreement with theoretical model calculations. Alpha particle heating has been unambiguously observed, its effect being separated successfully from possible isotope effects on energy confinement by varying the tritium concentration in otherwise similar discharges. The scan showed that there was no, or at most a very weak, isotope effect on the energy confinement time. The highest electron temperature was clearly correlated with the maximum alpha particle heating power and the optimum DT mixture; the maximum increase was 1.3±0.23 keV with 1.3 MW of alpha particle heating power, consistent with classical expectations for alpha particle confinement and heating. In the optimized shear regime, clear internal transport barriers were established for the first time in DT, with a power similar to that required in DD. The ion thermal conductivity in the plasma core approached neoclassical levels. Real time power control maintained the plasma core close to limits set by pressure gradient driven MHD instabilities, allowing 8.2 MW of DT fusion power with nDT(0)τEdiaTi(0) ≈ 1021 m-3 s keV, even though full optimization was not possible within the imposed neutron budget. In addition, quasi-steady-state discharges with simultaneous internal and edge transport barriers have been produced with high confinement and a fusion power of up to 7 MW; these double barrier discharges show a great potential for steady state operation.

Tritium recycling and retention in JET

Abstract JETs 1997 Deuterium Tritium Experiment (DTE1) allows a detailed study of hydrogenic isotope recycling and retention in a pumped divertor configuration relevant to ITER. There appear to be two distinct forms of retained tritium. (1) A dynamic inventory which controls the fueling behaviour of a single discharge, and in particular determines the isotopic composition. This is shown to be consistent with neutral particle implantation over the whole vessel surface area. (2) A continually growing inventory, which plays a small role in the particle balance of a single discharge, but ultimately dominates the hydrogenic inventory for an experimental campaign comprising thousands of pulses. This will be the dominant retention mechanism in long-pulse devices like ITER. The JET retention scaled-up to ITER proportions suggests that ITER may reach its tritium inventory limit in less than 100 pulses.

First mirrors in ITER: material choice and deposition prevention/cleaning techniques

We present here our recent results on the development and testing of the first mirrors for the divertor Thomson scattering diagnostics in ITER. The Thomson scattering system is based on several large-scale (tens of centimetres) mirrors that will be located in an area with extremely high (3?10%) concentration of contaminants (mainly hydrocarbons) and our main concern is to prevent deposition-induced loss of mirror reflectivity in the spectral range 1000?1064?nm. The suggested design of the mirrors?a high-reflective metal layer on a Si substrate with an oxide coating?combines highly stable optical characteristics under deposition-dominated conditions with excellent mechanical properties. For the mirror layer materials we consider Ag and Al allowing the possibility of sharing the Thomson scattering mirror collecting system with a laser-induced fluorescence system operating in the visible range. Neutron tests of the mirrors of this design are presented along with numerical simulation of radiation damage and transmutation of mirror materials. To provide active protection of the large-scale mirrors we use a number of deposition-mitigating techniques simultaneously. Two main techniques among them, plasma treatment and blowing-out, are considered in detail. The plasma conditions appropriate for mirror cleaning are determined from experiments using plasma-induced erosion/deposition in a CH4/H2 gas mixture. We also report data on the numerical simulation of plasma parameters of a capacitively-coupled discharge calculated using a commercial CFD-ACE code. A comparison of these data with the results for mirror testing under deuterium ion bombardment illustrates the possibility of using the capacitively-coupled discharge for in situ non-destructive deposition mitigation/cleaning.

Diagnostic experience during deuterium-tritium experiments in JET, techniques and measurements

Abstract During 1997 JET was operated for an extensive period using a D–T mixture (DTE1). Changes in the design and operation of diagnostic systems made over the years in preparation for this phase are described. A number of diagnostic techniques have been deployed to measure the deuterium and tritium content of the plasma during DTE1 and their results are compared. All diagnostics with a direct vacuum interface with the main vessel have been fitted with tritium compatible pumps and their exhaust gases have been re-routed to the active gas handling plant. All items on the torus which could lead to a significant leak in the event of failure, were required to have double containment. Therefore, all windows, and a majority of bellows and feedthroughs, were designed and installed with a double barrier. Heated fibre hoses were installed to transmit plasma light beyond the biological shield for spectroscopic purposes. Blind fibres and fibre loops were also installed to study the effects of higher neutron fluxes on these fibres. A radiation-hardened video camera was installed to monitor the plasma during the DTE1 discharges. Extra shielding was installed on a number of diagnostics to deal with the higher neutron fluxes during DTE1. The effect of neutron radiation on electronics in the Torus Hall was studied. During DTE1 the tritium fraction was measured at the edge and in the core using several diagnostic methods. High resolution Balmer α line spectroscopy gave a measurement typical of the plasma edge region. In the JET sub-divertor volume the tritium concentration of the neutral gas was measured using Balmer α spectroscopy of a Penning gauge discharge. Using Neutral Particle Analysis, the tritium concentration was measured typically in a zone 20–40 cm from the plasma edge. Local core measurements of the tritium fraction have been made using active Balmer α charge exchange spectroscopy. The error on this measurement is, however, large,∼30%. After the discharge the tritium fraction of the exhaust was measured using the exhaust monitoring system. Using short deuterium neutral injection pulses allowed neutron rate measurements of the tritium concentration in the core region. A further technique used the measured neutron rate and calculated neutron rate from other plasma parameters to determine the tritium concentration.

Physical aspects of divertor Thomson scattering implementation on ITER

This paper describes the challenges of Thomson Scattering implementation in the ITER divertor and evaluates the capability to satisfy project requirements related to the range of the measured electron temperature and density. A number of aspects of data interpretation are also discussed. Although this assessment and the proposed solutions are considered in terms of ITER compatibility, they may also be of some use in currently operating magnetic confinement devices.

The ITER divertor Thomson scattering system: engineering and advanced hardware solutions

A divertor Thomson scattering (TS) system being developed for ITER has incorporated proven solutions from currently available TS systems. On the other hand any ITER diagnostic has to operate in a hostile environment and very restricted access geometry. Therefore the operation in an environment of intensive stray light, plasma background radiation, the necessity meet the requirement using only a 20 mm gap between divertor cassettes for plasma diagnosis as well as to measure plasma temperatures as low as 1 eV severely constrain the divertor TS diagnostic design. The challenging solutions of this novel diagnostic system which has to ensure its steady performance and also the operability and maintenance are the focus of this report. One of the most demanding parts of the in-vessel diagnostic equipment development is the design assessment using different engineering analyses. The task definition and first results of thermal, e/m and seismic analyses are provided. The process of further improving of the design involves identification of susceptible areas and multiple iterations of the design, as needed. One of the key points for all Thomson scattering diagnostics are the laser capabilities. A high-performance and high-power laser system using a steady-state and high-repetitive mode Nd:YAG laser (2J, 50–100Hz, 3ns) has been developed. The reduced laser pulse duration matched with high-speed low-noise APD detector can be very important under high background light level. For diagnostics such as Thomson scattering and Raman spectroscopy, a high-degree of discrimination against stray light at the laser wavelength is required for successful detection of wavelength-shifted light from the laser-plasma interaction region. For this case of high stray light level, a triple grating polychromator characterized by high rejection and high transmission has been designed and developed. The novel polychromator design minimizes stray light while still maintaining a relatively high transmission.

The effect of divertor geometry on divertor and core plasma performance in JET

Abstract JET has completed a series of experiments in the Mk I and Mk IIA divertors on the effects of increased geometrical closure and target orientation. The potential benefits from closure were expected to be enhanced volumetric energy loss in the divertor (detachment), increased divertor neutral pressure for better pumping and He exhaust, and reduced main chamber neutral pressure for reduced sputtering. The expected effects on neutral pressures were observed. In ohmic and L-modes this led to detachment at lower upstream density and reduced density limits, in qualitative agreement with code calculations. The pumping speed was increased by about a factor of three. Zeff did not reduce, despite the reduced main chamber neutral pressure. In ELMy H-modes the effects of closure were less distinct, which may have been due in part to ELMs striking the upper surfaces of the divertor and main chamber limiting surfaces. The density limit and confinement quality were unaffected by changes in divertor geometry. Increasing triangularity increased the density limit, but also raised Zeff. Confinement was degraded by either deuterium puffing or nitrogen puffing. Detachment occurred at the inner target between ELMs, but not at the outer target until confinement was strongly degraded. Vertical target ELMy H-modes have thinner SOLs and lower midplane separatrix densities than those run on horizontal targets in Mk IIA. Given the JET observations on the lack of sensitivity of core plasma ELMy H-mode performance to divertor geometry, it appears appropriate to review the possibility of simpler, lower cost divertor options than the deep divertor design currently proposed for ITER.

Edge transport barrier in JET hot ion H modes

The effects of changing beam and plasma species on the edge transport barrier are investigated for ELM-free hot ion H mode discharges from the recent DT experiments on JET. The measured pressure at the top of the pedestal is higher for mixed deuterium and tritium and pure tritium plasmas over and above the level measured in pure deuterium plasmas at the same heating power. The pedestal pressure increases with beam tritium concentration for mixed deuterium-tritium beam injection into deuterium plasmas where the measured edge tritium concentration remains low. Alpha heating plays a significant role in the core of such plasmas, and the possible impact on the edge is discussed together with possible direct isotopic effects. Heuristic models for the transport barrier width are proposed, and used to explore a wider range of edge measurements including full power DD and DT pulses. This analysis supports the plasma current and mass dependence for a barrier width set by the orbit loss of either thermal or fast ions, though it does not unambiguously distinguish between them. The fast ion hypothesis could well account for some of the JET observations, though more theoretical work and direct experimental measurement would be required to confirm this. An ad hoc model for the power loss through the separatrix, Ploss ∝ nedge2 Zeff,edgeIp-1, is proposed based on neoclassical theory, a ballooning limit to the edge gradient and a barrier width set by the poloidal ion gyroradius. Such a model is compared with experimental data from JET. In particular, the model ascribes the systematic difference in loss power between the Mark I and Mark II divertors to the change in the measured Zeff. This change in Zeff is consistent with the observed change in impurity production, which is described in some detail, together with a possible explanation provided by the temperature dependence of chemical sputtering.

Thomson scattering diagnostic systems in ITER

Thomson scattering (TS) is a proven diagnostic technique that will be implemented in ITER in three independent systems. The Edge TS will measure electron temperature Te and electron density ne profiles at high resolution in the region with r/a>0.8 (with a the minor radius). The Core TS will cover the region r/a<0.85 and shall be able to measure electron temperatures up to 40 keV . The Divertor TS will observe a segment of the divertor plasma more than 700 mm long and is designed to detect Te as low as 0.3 eV . The Edge and Core systems are primary contributors to Te and ne profiles. Both are installed in equatorial port 10 and very close together with the toroidal distance between the two laser beams of less than 600 mm at the first wall (~ 6° toroidal separation), a characteristic that should allow to reliably match the two profiles in the region 0.8<r/a<0.85. Today almost every existing fusion machine has one or more TS systems installed, therefore substantial experience has been accumulated worldwide on practical methods for the optimization of the technique. However the ITER environment is imposing specific loads (e.g. gamma and neutron radiation, temperatures, disruption-induced stresses) and also access and reliability constraints that require new designs for many of the sub-systems. The challenges and the proposed solutions for all three TS systems are presented.

RF discharge for in situ mirror surface recovery in ITER

Almost all optical diagnostic systems in ITER will require the implementation of mirror recovery and protection systems. Plasma cleaning is considered to be the most promising technique for the removal of metal deposits from optical surfaces. The engineering and physical aspects of RF discharge application for continuous or periodic plasma treatment are discussed with a focus on implementation under ITER conditions. The ion flux parameters obtained in capacitively coupled (CC) RF discharge were measured in the mock-up of a plasma cleaning system. The uniformity of sputtering in CC RF discharge with and without a magnetic field was studied experimentally for the cylindrical discharge reactor geometry and compared with numerical simulations. The sharp increase in the sputtering rate resulting from the non-uniform radial distribution of the ion flux was observed near the electrode edges. The longitudinal magnetic field improves sputtering uniformity. It was demonstrated that Al/Al2O3 deposits can be removed in the Ne and D2 plasma of CC RF discharge but long-term exposition results in the degradation of the polycrystalline molybdenum mirror surface. The efficiency of Al sputtering in the atmosphere containing O2 and N2 fractions was studied in the D2/O2 and D2/N2 plasma of glow discharge. The addition of 2% of oxygen or nitrogen increases the sputtering yield by 3–4 times as compared with that in a nominally pure D2 discharge. The impact of metal deposits on the performance of diagnostic mirrors is discussed. It was shown that an ultrathin metallic film with a thickness as low as a few nm may cause a significant degradation of diagnostic mirrors with a transparent coating.