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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.

During 1997 JET operation with D-T plasmas, 35 g of tritium were introduced into the torus, mainly by gas puffing. It was found that during this period, the torus tritium inventory would accumulate at a rate of about 40% of the input. After tritium operation ceased, the experimental program continued with deuterium- and hydrogen-fuelled experiments, during which time the tritium inventory decreased to about 17% of the total input. Techniques aimed at detritiation of the torus included methods using deuterium gas (such as deuterium pulsing) which were used in the middle of the experimental campaign, and methods which could adversely affect the torus vacuum conditions (such as air purges) which were reserved for the period after the experimental campaign. Whilst it was found that the plasma tritium fraction could be reduced to below the 1% level in a few days, the tritium inventory reached a virtually steady level of about 6 g by the end of the campaign.

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.

Trace amounts of tritium gas have been injected in short puAs into JET ELMy H-modes with a wide range of deuterium gas-fuelling rates. Analysis of the subsequent time evolution of the neutron profile and extraction of the particle transport coeAcients have allowed us to distinguish between broad classes of mechanism which have been suggested as explanations for the H-mode density limit. The high penetration probability (20%) and rapid transport (sE) of fuel ions are shown to be only weakly influenced by strong gas fuelling ‐ hence mixture control is possible even when the total electron content is clamped. ” 1999 JET Joint Undertaking, published by Elsevier Science B.V. All rights reserved.

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.

JET's 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.

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.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

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 stepdown, 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 stepdown 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 stepdown, 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 stepdown 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 stepdown

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

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.

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...

The perpendicular x-ray emission up a to few MeV of runaway electrons has been measured in JET low-density ohmic discharges by means of the fast electron bremsstrahlung profile monitor. A diffusion model simulating the temporal evolution of the line-integrated x-ray signals is used to determine the runaway radial transport coefficient in the central region of the plasma ( for r/a

Helium 3 minority fuel ions have been observed to be accelerated to megaelectron-volt (MeV) energies by ion cyclotron radiofrequency (ICRF) heating. These energetic ions undergo nuclear reactions with impurity 9Be and 12C ions in the plasma, and characteristic gamma radiation is emitted. For special circumstances, this gamma ray emission has been detected with a 19 channel profile monitor constructed primarily for neutron measurements. Two-dimensional (2-D) profiles of the gamma radiation emission have been derived by tomographic reconstruction; these profiles correspond to a cross-section weighted density of 3He ions having energies of 2 or more MeV. As expected for ICRF heating, the observed spatial profile of the gamma radiation shows that access by the accelerated ions to the plasma volume on the high field side of the RF resonance layer is effectively inhibited. In addition, the profile appears to demand the presence of a group of passing particles that are localized in poloidal extent and are circulating some 15 cm to the low field side of the resonance layer. Apart from the 15 cm displacement, these features are well reproduced by a model that combines a 2-D bounce averaged Fokker-Planck calculation with an orbit-following code. A unique example of a gamma ray spatial profile captured at the time of a sawtooth crash shows that the profile is relatively unaffected apart from the sudden obliteration of the passing group, for which the ions are found to possess relatively high energy, presumably gained from the RF waves through the Doppler shift effect. On the basis of these observations, an explanation is proposed for the selective expulsion of RF heated fast ions from the central region of the plasma to the outer regions

A comprehensive set of fast particle diagnostics is routinely used at JET. Some are in the process of being upgraded and others, completely new, are being prepared for use during the forthcoming tritium experiments. For fusion product studies, the strength and profile of the charged particle birth distribution is obtained by measuring the neutron emission with three pairs of absolutely calibrated fission chambers and a two-camera profile monitor. Information on the DD neutron energy spectrum is deduced from a time-of-flight neutron spectrometer. Neutrons of 14 MeV energy from triton burnup are measured using silicon diodes and a high energy branch of the profile monitor. Absolute calibration is obtained with an activation system. A prototype lost alpha particle detector (Faraday cup) has been tested in the laboratory and has been installed inside JET so that its noise immunity can be tested. Two 14 MeV neutron spectrometers are under commissioning and a third is under construction. A high energy neutral particle analyser is routinely used to diagnose fast RF driven particles and will also be available for alpha particle studies. The intensity of RF driven fast particles is also deduced by the spectroscopy of gamma rays emitted in reactions with impurity ions in the plasma. The gamma ray measuring branch of the neutron profile monitor adds spatial resolution to the measurement. A new antenna for measuring ion cyclotron emission (ICE) and a gyrotron for alpha particle scattering experiments are in the process of being commissioned. Active charge exchange spectroscopy to investigate the low energy range of the alpha particle population is being considered. Where appropriate, information gathered with these systems is presented to illustrate their performance

Pellet enhanced performance (PEP) has been observed in a number of JET discharges at various plasma conditions, in both L and H modes, with the H multiplier (the confinement enhancement factor over the Goldston confinement time) covering the range from 1 to 4, and with plasma currents from 1 MA to 4.1 MA. Most of the PEP plasmas have been created by refuelling with pellets of 4 mm diameter injected at 1.2 km/s. PEPs show an improved central confinement with an effective heat conductivity reduced by factors of approximately 2-5 relative to otherwise comparable discharges. This is possibly related to the inverted shear in the plasma core due to the large local bootstrap current density. The limitations in the PEP performance seem to be set by at least two mechanisms: impurity behaviour, MHD activity or a combination of both. In certain discharges, MHD modes seem to be able to check the often observed impurity accumulation. Too much MHD mode activity, however, easily destroys the enhanced confinement of the PEP discharge. The stability of the ballooning modes has been studied and the PEP plasma core is found to be in the second stability region against ballooning modes or close to marginal stability. In a number of discharges complex high (m,n) modes have been observed with the soft X-ray cameras. The behaviour of the low (m,n) MHD modes can only be understood by considering the detailed evolution of the inverted q profile, which exists in a given discharge

The JET neutron emission profile monitor is used to measure the 2.5 MeV and 14 MeV neutron emission line integrals before and after sawtooth crashes in high d-d neutron yield, hot ion H mode plasmas in the Joint European Torus (JET). Deuterium-deuterium (d-d) fusion produces 2.5 MeV neutrons and 1 MeV tritons (t) at nearly equal rates from its two reaction channels. A plasma current of 3 MA is sufficiently high to contain most of the fusion product tritons, which have birth orbit gyroradii and velocity space distributions similar to those of the 3.5 MeV or particles from d-t fusion. By examining neutron emission line integrals and tomographically deduced local emissivity profiles, an upper limit of 10% can be placed on the net fraction of fusion product tritons which are displaced from the plasma axis by those sawtooth crashes studied. This is a much smaller net fraction than that typically observed, 35-55%, for displaced injected neutral beam deuterium ions. A study of the response of beam injected deuterium ions to a sawtooth crash shows that the change in their axial density depends on the precrash spatial width of the neutron emissivity profile. The fusion product tritons, which have a large precrash spatial width, respond weakly to a crash. This weak response is consistent with the behaviour of the analogous d-d beam thermal neutrons when extrapolated to the corresponding emissivity spatial width. The implication of these observations is that beam ions and 3.5 MeV alpha particles in JET may be relatively resilient to sawtooth clashes, when the spatial width of their density is sufficiently large

Spatial profiles of the neutron emission from deuterium plasmas are routinely obtained at the Joint European Torus (JET) using the line-integrated signals measured with a multichannel instrument. It is shown that the manner in which these profiles relax following the termination of strong heating with neutral beam injection (NBI) permits the local thermal diffusivity ( chi i) to be obtained with an accuracy of about 20%. The profiles of chi i for minor radii in the range 0.25

The JET neutron emission profile monitor was used to study ohmically heated deuterium discharges. The radial profile of the neutron emissivity is deduced from the line-integral data. The profiles of ion temperature, Ti, and ion thermal diffusivity, chi i, are derived under steady-state conditions. The inferred chi i, for 0.3 < p = r/a < 0.5, is in the range 0.5-2.5m2s-1. The ion thermal diffusivity is higher than, and its scaling with plasma current opposite to, that predicted by neoclassical theory.

During a series of experiments with tritium (T) in deuterium (D) plasmas in the Joint European Torus (JET), the temporal evolution and the two dimensional (2-D) spatial profiles of the 2.5 and 14 MeV neutron emissivities from D-D and D-T fusion reactions were inferred from measurements with the JET neutron emission profile monitor. These experiments, involving triton production from D-D fusion, beam deposition and diffusion, D-T fusion, and tritium removed from wall tiles, were investigated in four plasma scenarios: (i) In high performance deuterium plasmas with deuterium neutral beam injection, the 14 MeV neutron emissivity due to triton burnup was observed. (ii) In discharges with 1% tritium beam injection, neutron emissivity ratios showed that approximately the same deposition profiles resulted from tritium as from deuterium beams. A thermalized tritium diffusion experiment was performed in which the T-D density ratio was found to be spatially constant across the plasma; in conjunction with similar particle source profiles, this indicates that deuterium and tritium have similar particle transport properties. (iii) In two high performance discharges for which two of the sixteen neutral beam sources operated with 100% tritium, the production rate of 14 MeV neutrons reached 6 × 1017 n.s-1 The alpha particle 2-D birth profile was directly inferred from the measured 14 MeV neutron emissivity profile. Both the axial 14 MeV neutron emissivity and the axial ion temperature saturated before the maximum global emission was reached. (iv) During the tritium cleanup phase, residual tritium entering the plasma produced a spatially constant ratio of tritium to deuterium, confirming the similarity of their particle transport properties

Preliminary experiments were carried out in the Joint European Torus (JET) using 120 keV helium (He) neutral beam injection (NBI). Injected power levels up to 5 MW with 3He and 7 MW with 4He, lasting up to 3 s were reached. The 3 s helium NBI produced efficient ion heating and similar global and local energy confinements to those obtained with deuterium (D) NBI, in both limiter and X-point plasma geometries, in L-mode and H-mode plasma regimes. The elimination of beam-plasma and beam-beam fusion reactions by replacing D NBI with He NBI extended the range for measuring ion temperatures with the JET neutron profile monitor and neutron spectrometers.

Measurements from the JET neutron profile monitor are analyzed tomographically to deduce the 2-D spatial distribution of neutron emissivity during NBI heating both before and after sawtooth crashes. In a discharge where the global neutron emission decreases slightly to 5/6 of the pre-crash rate of nearly 1016 n s-1, the axial emissivity drops much more to only 1/6 of the pre-crash value. The sources of neutron emissivity are analyzed with an analytic Fokker-Planck formulation. The main change in neutron emissivity is due to fast ion redistribution during a sawtooth crash and a decrease in beam-beam-produced emissivity. The deduced fast ion redistribution is consistent with other observations of sawtooth behaviour in JET.

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.

There is evidence in JET (Joint European Torus) and in other tokamak machines that large currents flow to the vessel walls during vertical instabilities and the subsequent current quench. The extent and the location of damage to the first wall components in JET caused by the interaction of the magnetic field with currents through these components are reported. An analysis of the damage allows one to estimate the currents. The difference in the toroidal field measured at the top and the bottom Inside the vessel gives an indication of the total poloidal current flowing between the plasma and the vessel. The voltage difference along 2.25 m long poloidal sections at the top and bottom of the vessel gives an indication of the net repelling (stabilizing) force between the plasma and the vessel. In addition, current shunts on strategically located tiles in the vessel were used to measure local current flowing across the tile supports. The estimates are compared, and the direction of the forces exerted on the in-vessel components and the implications that these currents have for the design of the new JET pumped divertor are discussed. >

This paper presents the reference design of the radially viewing horizontal neutron profile monitor for ITER. It incorporates integral neutron shielding, and with full machine integration, should provide control and plasma diagnostic information based on the deduced neutron emission profile. Information can be obtained on total neutron emission (fuel burn-up rate), alpha-particle birth profile, plasma position and the time dependent effects of plasma instabilities such as sawteeth.

Publisher Summary 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.

High performance plasmas were obtained in many JET experiments, both before and after the major shut down of 1992–93. In this paper, a summary is presented, of the MHD activity which, in both the old and the new JET configuration, appears to be most clearly correlated to the collapse of the neutron yield. The paper will discuss aspects of MHD observations in three types of performance limitation; a) slow roll‐overs of the neutron rate, which appear to start with the growth of low n modes in the outer 30% of the plasma, b) sawteeth coupled to ELMs, and c) giant ELMs.

Various studies have shown that a significant improvement in tokamak performance can be obtained by elongating and shaping the cross section of the plasma ring. Plasmas with high elongation are however strongly unstable to the vertical axisymmetric mode and their control is a major problem. The objective of TCV (Tokamak a Configuration Variable Ro = 0.88 m, a = 0.24 m, b/a up to 3, Btor = 1.43 T, Iplasma up to 1.2 MA) presently under construction at the CRPP is to study scenarios for the creation of such plasmas and to evaluate the improvements in plasma parameters such as the maximum plasma current and density, the confinement time and the maximum pressure that can be sustained.

A steady-state current can be produced in a toroidal plasma by absorption of an electron cyclotron heating wave in the region of a localized magnetic field gradient along the magnetic field lines. The current produced by this mechanism flows in the same direction as the Fisch-Boozer current and it is equivalent in magnitude for reasonable values of ΔB/B. The scaling of the ratio current to RF power absorbed with the plasma variables n, T, and R (major radius), is the same for both currents, namely T/nR.

Experiments are described in which a 2-mm-diameter cylindrical pellet of solid hydrogen is injected at low velocity to the centre of the upper half of the Doublet III tokamak, immediately before initiation of the plasma discharge. The pellet ablates in 6–9 ms and increases the initial plasma density by a factor of 2 to 3. The profiles of both electron density and current density of discharges initiated with a pellet are very flat compared to normally initiated discharges, and lower electron temperatures and lower impurity levels persist for hundreds of milliseconds. The resultant dee-shaped plasmas have height-to-width ratios of up to 1.8 on axis and 2.0 on the outermost flux surface, and are suitable targets for neutral-beam or electron cyclotron heating.

Using a neutral-beam injection power of 3.4 M W, volume-averaged toroidal betas of up to βT = 4.5% have been obtained in low-toroidal-field, low-qψ, vertically elongated discharges in the Doublet III tokamak. This level of βT is above the minimum level required for a tokamak reactor, thus demonstrating that reactor level values of βT are possible in a tokamak device. The observed enhancement of βT with vertical elongation lends confidence in the design of future devices which rely on vertical elongation.

Large-amplitude sawtooth oscillations have been observed in the soft-X-ray emission, central electron temperature, neutron production rate, radiated power, Dα emission, fast-neutral flux and one-turn voltage during beam-heating experiments in the Doublet III tokamak. The necessary conditions for the appearance of giant sawteeth seem to be operation at low qa and high βT. Many features of the giant sawteeth have been simulated with a transport code to which a sawtooth mixing model and several diagnostic models were added. Substantial redistribution of particles and energy evidently occurs over the plasma interior. Global energy confinement, however, is only moderately degraded by the sawteeth.

It is demonstrated numerically that a tokamak plasma can be evolved continuously from a near-circular cross-section shape to a 4/1 vertically elongated racetrack. All intermediate stages and the final state are stable to axisymmetric magnetohydrodynamic modes. The stabilization is provided by the vacuum vessel walls on the ideal time scale and by an orthogonal active feedback system on the resistive time scale.

Analysis of the expanded boundary divertor in Doublet III reveals a medium density operating regime ( n e ≲ 4 × 10 13 cm -3 ), in which (1) the divertor density ne(Div) rises with the cube of the main plasma line average density, n e , and (2) the neutral pressure at the exhaust end of the divertor has a quadratic dependence on n e (with divertor plasma temperatures ≲ 15 eV). In this regime, injection of argon into the discharge results in a steady state concentration of argon, which is two orders of magnitude higher in the divertor plasma than in the main discharge. This trapping effect suggests that the argon concentration in the divertor region may be useful as an externally controllable parameter; that is, injected argon effectively cools the plasma boundary, yet is not deleterious to the plasma as a whole. A model which successfully describes the behavior of the edge and divertor plasmas in ohmically heated discharges suggests that these desirable trapping and radiative features may be maintained in the high powered neutral beam heating phase.

Detailed scrape-off measurements during Alfven wave heating experiments ( P RF ~ 200 kW) in TCA are presented. The Measurements are performed by means of electrical probes and by deposition probes fixed onto the antennae. It is shown that different antenna designs strongly influence the behaviour of the boundary plasma. Even during low power RF heating experiments is it found that the plasma-antenna interaction leads to a non-Maxwellian plasma. Low frequency density fluctuations are associated with changes of the density and temperature profiles during RF. High frequency fluctuations at the driving frequency and their harmonics are observed.

This report describes the experimental measurements and data analysis techniques used to evaluate the energy confinement in noncircular plasmas produced in Doublet III. Major aspects of the confinement measurements and analysis techniques are summarized. Machine parameters, diagnostic systems and discharge parameters relavent to the confinement measurements are given. Magnetic analysis techniques used to determine the plasma shape are reviewed. Scaling of the on-axis values of electron temperature, confinement time and Z/sub eff/ with plasma density is presented. Comparison with scaling results from other circular tokamaks is discussed. Numerical and analytic techniques developed for calculating the plasma energy confinement time and self-consistent profiles of density, temperature, current, and flux in non-circular geometries are described. These techniques are applied to the data and used to determine the central and global electron energy confinement time for a typical doublet plasma. Additional aspects of the confinement such as the radial dependence of the electron thermal conductivity and the estimated ion temperature are explored with the aid of a non-circular transport simulation code. The results of the confinement measurements are summarized and discussed. A brief summary of the theoretically expected effects of noncircularity on plasma confinement is included for reference as Appendix I.

Plasma Engineering includes both basic investigations and the application of research results to fusion system design. Plasma Engineering provides fusion physics input to various CTR programs so that the effects of system operating limitations and requirements can be considered in the design process. In this report a broad scope of Plasma Engineering information is presented. For example, tokamak dimensions and scaling, magnetic field requirements, impurity control and wall material criteria, poloidal field system design, effects of changing the number and shape of the toroidal field coils, plasma ionization and breakdown, fueling, plasma heating by neutral beam injection, vacuum requirements and components, low-neutron-yield testing and prospects for clean fuels are discussed. The analyses and techniques developed are specifically applied to the Oak Ridge Experimental Power Reactor (EPR).

Summary The first Neutral Injection System Beamline became operational on the Doublet III tokamak in September 1981. Power delivery tests and system debugging proceeded through December and initial experiments on plasma heating were started in January 1982. This first beamline has been running well and as of February 1982 is routinely delivering about 2.0 MW of neutral hydrogen power to various circular and elongated dee-shaped deuterium plasmas. The second beamline is now being brought on-line and we expect to have 5–6 MW of heating power available for definitive β studies in mid 1982. We find that the change in the total plasma stored energy caused by the beams is essentially independent of toroidal field and plasma current, but increases weakly with density. The total stored energy increases with current and density, but is independent of toroidal field. A volume-average βT= 2.2% is obtained in low toroidal field discharges.

A phased program of plasma heating at the electron cyclotron frequency is proposed for the Oak Ridge tokamaks ISX and ORMAK Upgrade. The past history of the program of electron cyclotron heating (ECH) at ORNL on mirrors and in the ELMO Bumpy Torus has been successful. Future technological developments in the production of high power high frequency microwave tubes look promising at this time. The physics of wave propagation and particle heating are fairly well understood and indicate the viability of this technique. Studies on breakdown and on runaway electron reduction will provide useful information for larger machines. Recent experiments in the USSR on small tokamaks have shown that ECH is a viable heating technique. Providing that the microwave tubes become available, the engineering considerations suggest that the technique is practical and workable, based on present day technology.

At the present state of knowledge, the design of fusion reactors is based on scaling which has not been fully verified experimentally. Various choices of scaling laws can be made, leading to different results. Design studies demonstrate sensitivity to the scaling laws assumed. Models are described suitable for simulating the energy and particle balances of a tokamak plasma, for surveying the range of parameters possible during operation, and for surveying the sensitivities of the parameters to system changes. Physical models for transport losses, beam-plasma interactions, noncircular plasma behavior, sputtering, impurity effects, and reactor control are included. The calculations based on these models are tested against present experimental results and are used to predict the behavior of future devices. Power balance results and their sensitivities are focused primarily on the ORNL Experimental Power Reactor (EPR), Demonstration Power Reactor (Demo), and Commercial Power Reactor (CPR) Designs. The sizes of these plants and the ranges of their possible power outputs are discussed. The limitations on the operation of these devices due to constraints such as beta limits, sputtering, and impurity buildup are described. Conclusions which result from applying the model to the systems noted above, but which are applicable to tokamak design in general, are presented.

Growth rates of axisymmetric modes of elongated tokamak plasmas, in the presence of a conducting shell, are computed, and the effects of plasma-wall distance, current profile and elongation are analysed. For highly elongated plasmas, it is found that the side walls are more important than the top and bottom walls for stabilizing these modes. Furthermore, it is shown that decreasing the width of the radial current profile has a much greater effect on vertical stability than increasing the width of the conducting shell by the same amount. Finally, the conditions for axisymmetric stability during the evolution from a circular plasma to a highly elongated racetrack are investigated.

The particle and energy transport properties of the high fusion performance JET pulses that were obtained before and during the first tritium experiments are discussed. The particle diffusion coefficient of tritium is determined by monitoring the decay of a small quantity of injected tritium in a deuterium background plasma. A good simulation of the measured 14 MeV neutron emissivity profile is obtained throughout the decay phase if the mixing of the two species is described by a model in which the tritium diffusion coefficient is similar to that of deuterium. The energy confinement of these low density, hot ion, H mode discharges is found to have both improved central and edge confinement over the conventional medium to high density H mode discharges, regardless of the presence or absence of tritium in the discharge. As the tritium concentration of these D-T discharges is small (varying from

The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Simulations show that pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations in addition to equilibrium changes. It is demonstrated that these perturbations may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. This method works because coils inside the vessel do not have a flux penetration delay. Vertical disruption induced voltages and forces on the rapid coils and the vessel are investigated to establish coil and vessel design requirements. A model is proposed and used to estimate the additional vertical force due to poloidal currents, which is comparable to the force from toroidal currents.