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.

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JET tokamak using H mode divertor discharges with ITER-like shapes and safety factors. Deuterium minority heating in tritium plasmas was demonstrated for the first time. For 9% deuterium, an ICRH power of 6 MW gave 1.66 MW of fusion power from reactions between suprathermal deuterons and thermal tritons. The Q value of the steady state discharge reached 0.22 for the length of the RF flat-top (2.7 s), corresponding to three plasma energy replacement times. The Doppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimum for fusion and close to the critical energy. Thus, strong bulk ion heating was obtained at the same time as high fusion efficiency. Deuterium fractions around 20% produced the strongest ion heating together with a strong reduction of the suprathermal deuteron tail. The ELMs had low amplitude and high frequency and each ELM transported less plasma energy content than the 1% required by ITER. The energy confinement time, on the ITERH97-P scale, was 0.90, which is sufficient for ignition in ITER. 3He minority heating, in approximately 50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating. Central ion temperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV. The normalized H mode confinement time was 0.95. Second harmonic tritium heating produced energetic tritons above the critical energy. This scheme heats the electrons in JET, unlike in ITER where the lower power density will allow mainly ion heating. The inverted scenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successful heating method producing both suprathermal neutrons and bulk ion heating. Theoretical calculations of the DT reactivity mostly give excellent agreement with the measured reaction rates.

High power combined NBI + ICRF heating experiments have been carried out in the JET Mark IIa divertor configuration in the hot ion ELM-free H mode regime, both in deuterium (DD) and in deuterium-tritium (DT) plasmas. Results are presented from a wide range of additional heating power levels, ICRF up to 9.5 MW tuned to the fundamental hydrogen minority, NBI up to 22 MW, and for plasma currents up to 4.2 MA and toroidal fields up to 3.6 T. Discharges with combined NBI + ICRF heating show a clear improvement in electron temperature, DD neutron yield and stored energy with respect to NBI only discharges. High energy neutral particle analyser data show that acceleration of the NBI deuterons takes place due to absorption of ICRF power at the second harmonic deuterium resonance. This is confirmed by numerical simulations with the PION code, indicating that up to 40% of the ICRF power is absorbed by bulk and NBI ions. ICRF heating has been an essential ingredient in the DT experiments in the ELM-free hot ion regime, contributing to the achievement of a record fusion power of 16.1 MW and a record stored energy of 17 MJ.

Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This article analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, we examine proposed compact cameras, in which neutron fluxes are inferred from N16 decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with N16-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins.

The JET neutron profile monitor was used to study the transport of tritium into a magnetically confined deuterium plasma. Trace amounts of tritium were introduced through a gas valve beyond the plasma edge into a variety of plasma confinement regimes. The d(t,n)α fusion cross section is two orders of magnitude greater than the d(d,n)3He cross section and so a puff of tritium which has a negligible effect on the plasma nevertheless produces a large d–t neutron signal. The profile monitor consists of two cameras each made up of a fan-shaped array of collimated lines of sight. It was used to measure the d–d and d–t neutron profiles simultaneously. This article describes the detection system, its operation and assesses the difficulties due to scattered neutrons. The profiles can be used to determine tritium density and transport coefficients.

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 Alfvénic 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. © 1999, Euratom

The performance of diagnostic systems are limited by their weakest constituents, including their calibration issues. Neutron diagnostics are notorious for problems encountered while determining their absolute calibrations, due mainly to the nature of the neutron transport problem. In order to facilitate the determination of an accurate and precise calibration, the diagnostic design should be such as to minimize the scattered neutron flux. ITER will use a comprehensive set of neutron diagnostics--comprising radial and vertical neutron cameras, neutron spectrometers, a neutron activation system and internal and external fission chambers--to provide accurate measurements of fusion power and power densities as a function of time. The calibration of such an important diagnostic system merits careful consideration. Some thoughts have already been given to this subject during the conceptual design phase in relation to the time-integrated neutron activation and time-dependent neutron yield monitors. However, no overall calibration strategy has been worked out so far. This paper represents a first attempt to address this vital issue. Experience gained from present large tokamaks (JET, TFTR and JT60U) and proposals for ITER are reviewed. The need to use a 14-MeV neutron generator as opposed to radioactive sources for in-situ calibration of D-T diagnostics will be stressed. It is clear that the overall absolute determination of fusion power will have to rely on a combination of nuclear measuring techniques, for which the provision of accurate and independent calibrations will constitute an ongoing process as ITER moves from one phase of operation to the next.

Associacao EURATOM/IST, Centro de Fusao Nuclear, 1096 Lisbon, CODEX, Portugal1. IntroductionDuring the MKII Divertor campaign the performance characteristics of the Hot-Ion ELM-freeH-mode regime in Deuterium plasmas have been comprehensively explored [1,2].Experiments were carried out up to a plasma current of 4.2 MA and a toroidal field of 3.8 T,with Neutral Beam (NB) heating alone and with NB plus Hydrogen minority Ion CyclotronResonance Heating (ICRH), up to a total input power of about 25 MW.The maximum fusion reactivity is obtained transiently; density and stored energy risecontinuously with time, until the terminating MHD event, which is either a large sawtooth, aGiant ELM or an external kink mode. The identification of the so-called Outer Mode as anideal external kink, driven by the edge current gradient, led to the development of a methodfor its mitigation or avoidance by decreasing the plasma current during the ELM-free phase.As a result the performance, robustness and reliability of the regime have been improved,extending the maximum DD neutron yield to 5.2×10

Ion cyclotron resonance heating (ICRH) experiments have been carried out in JET D-T plasmas using scenarios applicable to reactors. Deuterium minority heating in tritium plasmas is used for the first time and produces 1.66 MW of D-T fusion power for an ICRH power of 6 MW. The Q value is 0.22, which is a record for steady state discharges. Fundamental He-3 minority ICRH, in both 50:50 D-T and tritium dominated plasmas, generates strong bulk ion heating and ion temperatures up to 13 keV. Second harmonic tritium ICRH is seen to heat mainly the electrons as expected for JET conditions. All three schemes produce H-mode plasmas.

In nearly all hot ion H modes in JET, a magnetohydrodynamic (MHD) event is clearly observed just before the time the stored plasma energy saturates and the neutron yield starts to decline. The results of a systematic analysis of MHD observations for a large number of discharges is reported. The relationship between MHD phenomena and the onset of confinement limitation is discussed, as are aspects of the three main types of performance limiting MHD: (a) low-n modes in the outer regions of the plasma, (b) sawteeth and (c) giant edge localized modes (ELMs). Model simulations indicate how the transport is affected and allow an assessment of how much the neutron yield would be improved if the MHD activity were absent

Based upon JET experience, the measurement of multichannel collimated neutron fluxes and of the neutron spectrum gives time-dependent information on spatial profiles of neutron emission and alpha–particle birth, the total neutron emission (fuel burn-up rate), plasma position, effects of plasma instabilities, triton burn-up, ion temperature, and fuel densities. The design for a horizontally viewing neutron camera for ITER is based upon the prototype and upgrade versions of the JET neutron emission profile monitor and the JET spectrometers. It is proposed that vertically stacked modules are installed in the ITER biological shield in a fan shaped viewing geometry, aimed at a focal point located at the slit opening of a preshield designed to reduce the streaming neutron flux. Each module contains a pair of sight lines with adjustable collimation, allowing for multiple detector neutron flux monitoring and neutron spectroscopy over a wide operating range. The modular system allows flexibility in detector choice and viewing geometry.

The prototype neutron profile monitor was in use at the Joint European Torus (JET) from 1987 until the end of 1994, when it was upgraded to incorporate features which experience showed to be desirable. For ohmic and ion cyclotron resonance frequency (ICRF)-heated discharges, the profile monitor provides useful information on ion temperatures and deuterium densities. However, most JET discharges involve beam heating, for which charge exchange recombination spectroscopy provides high quality ion temperature profile data, so that analysis of neutron profile monitor data has been diverted to other matters, such as the following: the examination of the effects of beam deposition and density profiles on the neutron emissivity profile; the extraction of information on fast particle diffusion from a study of the bum-up of D-D reaction fusion product tritons; the interpretation of results from the first tritium experiment; studies of gamma-ray emission from ICRF-accelerated fast ions interacting with plasma impurities; and the interpretation of sawtooth crash data. The profile monitor has also been used to measure bremsstrahlung from fast electrons generated using a lower hybrid current drive. The requirements which led to the upgrading of the prototype diagnostic are indicated and anticipated future applications are summarized.

The remote maintenance of a tokamak type fusion experiment requires a system to satisfy unscheduled repair after failure, scheduled replacement of parts due to consumable depletion or erosion and modification of the tokamak in accordance with experimental requirements. The preparation of remote handling equipment includes the design and manufacture of new equipment, proving of equipment reliability, development and proving of operating procedures, training of operators and setting up an operations support system. The remote handling system is required to be prepared in a time frame dictated by the lead time of the components to be handled. The remote handling philosophy has a direct influence on the design of remote handling equipment, the design of tokamak components, and the remote handling operations. The principle of component handling using teleoperation is characterised as the man-in-the-loop. The handling task is performed by an operator using a manipulator which provides force, and visual and audio feedback of sufficient resolution to create the atmosphere for the operator of actually being in the task environment.

A power step-down approach leads to extended, high performance operation of deuterium plasmas in the hot ion edge localized mode (ELM)-free H mode of the JET tokamak. ELM-free discharges in JET, when heated steadily at the maximum power available, are usually transient and terminate with a variety of MHD phenomena. With power step-down, instabilities are delayed or avoided, and nearly constant plasma conditions at up to 10 MJ of stored energy are maintained by 10 MW of heating power for up to 1 s, about an energy confinement time, at 3.5 MA and 3.4 T. No large transient corrections are required for confinement calculations in this hot ion regime, and the energy confinement time scaling is found to be similar to global ELM-free H mode confinement scaling. The confinement times can increase after the transition to quasi-steady conditions, to some 10-25% above the scaling law predictions. Code simulations of the experimentally observed neutron rate and the lack of a major discontinuity in this rate before and after beam power step-down confirm that neutron production is predominantly (>60%) from thermal fusion. The scaling of fusion rates from deuterium plasmas to deuterium-tritium (DT) mixtures yields a fusion Q approaching 1. After power step-down, the edge pressure gradient stops increasing and helps delay the onset of instabilities. Giant ELMs still occur, associated with 'outer mode' activity and continuously rising plasma density. The plasma remains in an ELM-free H mode even when the heating power after power step-down is lower than the L-H transition power. Extended operation optimization experiments indicate that stored energy and neutron production are maximized and the density rise minimized by preferentially injecting high energy 140 keV beams after power step-down.