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The effect of isotope on confinement in high‐recycling, L‐mode plasmas is studied on the Tokamak Fusion Test Reactor (TFTR) [see D. M. Meade, J. Fusion Energy 7, 107 (1988)] by comparing hydrogen and deuterium plasmas with the same magnetic field and similar electron densities and heating power, with both Ohmic and deuterium‐neutral‐beam heating. Following a long operational period in deuterium, nominally hydrogen plasmas were created through hydrogen glow discharge and hydrogen gas puffing in Ohmic plasmas, which saturated the exposed limiter surface with hydrogen and raised the H/(H+D) ratio from 10±3% to 65±5%. Ohmic deuterium discharges obtained higher stored energy and lower loop voltage than hydrogen discharges with similar limiter conditions. Neutral‐beam power scans were conducted in L‐mode plasmas at minor radii of 50 and 80 cm, with plasma currents of 0.7 and 1.4 MA. To minimize transport differences from the beam deposition profile and beam heating, deuterium neutral beams were used to heat the...

The TRANSP code [R. V. Budny et al., Nucl. Fusion 35, 1497 (1995)] is used to construct comprehensive, self‐consistent models for plasmas within the separatrix surface in the International Thermonuclear Experimental Reactor (ITER) [Technical Basis for the ITER Interim Design Report, Cost Review and Safety Analysis (International Atomic Energy Agency, Vienna, 1996)]. Steady state profiles of two plasmas from the ITER ‘‘Interim Design’’ database are used. Effects of 1 MeV neutral beam injection, sawteeth mixing, toroidal field ripple, and helium ash transport are included. Results are given for the fusion rate profiles, and parameters describing effects such as the alpha particle heating of electrons and thermal ions, and the thermalization rates. The modeling indicates that the deposition of the neutral beam ions will peak in the plasma center, and the average beam ion energy will be half the injected energy. Sawtooth mixing will broaden the fast alpha profile. The toroidal ripple loss rate of alpha energy...

High or enhanced confinement (H‐mode) plasmas have been obtained for the first time with nearly equal concentrations of deuterium and tritium in high‐temperature, high poloidal beta plasmas in the Tokamak Fusion Test Reactor (TFTR) [McGuire, Phys. Plasmas 2, 2176 (1995)]. Tritium fueling was provided mainly through high‐power neutral beam injection (NBI) with powers up to 31 MW and beam energies of 90–110 keV. A transition to a circular limiter H‐mode configuration has been obtained, following a programmed rapid decrease of the plasma current. Isotope effects, due to the presence of tritium, led to different behavior for deuterium–deuterium (DD) and deuterium–tritium (DT) H‐modes relative to confinement, edge localized magnetohydrodynamic modes (ELMs), and ELM effects on fusion products. However, the threshold power for the H‐mode transition was the same in DD and DT. Some of the highest values of the global energy confinement time, τE, have been achieved on TFTR during the ELM‐free phase of DT H‐mode plasmas. Enhancements of τE greater than four times the L‐mode have been attained.

The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in the Tokamak Fusion Test Reactor [Plasma Phys. Controlled Fusion 26, 11 (1984)]. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. Radial profiles of the deuterium and tritium densities are determined from the D–T fusion neutron emission profile. The inferred scalings with average isotopic mass are quite strong, with τE∝〈A〉0.85±0.20, τEthermal∝〈A〉0.89±0.20, χitot∝〈A〉−2.6±0.5, and De∝〈A〉−1.4±0.2 at fixed Pinj. For fixed local plasma parameters χitot∝〈A〉−1.8±0.4 is obtained. The quoted 2σ uncertainties include contributions from both diagnostic errors and shot irreproducibility, and are conservatively constructed to attribute the entire scatter in the regressed parameters to uncertainties in the exponent on plasma mass.

An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium–tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR) [Proceedings of the 14th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, 1992 (International Atomic Energy Agency Vienna, 1992)]. A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the transp [‘‘A standard D–T supershot simulation’’ (to be published in Nucl. Fusion, 1994)] code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing‐down path due to reduction in banana thickness, which enables one to understand, for the first time, the dependence of the collisional loss rate on Zeff.

In the Tokamak Fusion Test Reactor (TFTR) [International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, Paper No. A‐2‐2 (International Atomic Energy Agency, Vienna, 1993)] there have been at least three types of anomalous loss of alpha‐like deuterium–deuterium (D–D) fusion products: (1) a magnetohydrodynamic (MHD)‐induced loss of D–D fusion products correlated with Mirnov and fishbone‐type oscillations and sawtooth crashes, (2) a slow ‘‘delayed’’ loss of partially thermalized D–D fusion products occurring without large MHD activity, and (3) ion cyclotron resonance heating (ICRH)‐induced loss of D–D fusion products ions observed during direct electron heating experiments, and possibly also during 3He minority heating. In this paper each of these will be reviewed, concentrating on those due to MHD activity, which are the largest of these anomalous losses. The experimental results are compared with numerical models of various fusion product transport mechanisms.

Using perturbations in electron density and temperature induced by small helium gas puffs in TFTR (Tokamak Fusion Test Reactor) [Plasma Phys. Controlled Nucl. Fusion Res. 1, 51 (1986)], the dominance of convective heat transport in the core (r/a

The complete ion cyclotron range of frequency (ICRF) heating system for the Tokamak Fusion Test Reactor (TFTR) [Fusion Tech. 21, 1324 (1992)], consisting of four antennas and six generators designed to deliver 12.5 MW to the TFTR plasma, has now been installed. Recently a series of experiments has been conducted to explore the effect of ICRF heating on the performance of low recycling, supershot plasmas in minority and nonresonant electron heating regimes. The addition of up to 7.4 MW of ICRF power to full size (R∼2.6 m, a∼0.95 m), helium‐3 minority, deuterium supershots heated with up to 30 MW of deuterium neutral‐beam injection has resulted in a significant increase in core electron temperature (ΔTe=3–4 keV). Simulations of equivalent deuterium–tritium (D–T) supershots predict that such ICRF heating should result in an increase in βα(0)∼30%. Direct electron heating has been observed and has been found to be in agreement with theory. The ICRF heating has also been coupled to neutral‐beam heated plasmas f...

General plasma physics principles state that power flow Q(r) through a magnetic surface in a tokamak should scale as Q(r)= {32π2Rr3Te2c nea/[eB (a2−r2)2]} F(ρ*,β,ν*,r/a,q,s,r/R,...) where the arguments of F are local, nondimensional plasma parameters and nondimensional gradients. This paper reports an experimental determination of how F varies with normalized gyroradius ρ*≡(2TeMi)1/2c/eBa and collisionality ν*≡(R/r)3/2qRνe(me/ 2Te)1/2 for discharges prepared so that other nondimensional parameters remain close to constant. Tokamak Fusion Test Reactor (TFTR) [D. M. Meade et al., in Plasma Physics and Controlled Nuclear Fusion Research, 1990, Proceedings of the 13th International Conference, Washington (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] L‐mode data show F to be independent of ρ* and numerically small, corresponding to Bohm scaling with a small multiplicative constant. By contrast, most theories predict gyro‐Bohm scaling: F∝ρ*. Bohm scaling implies that the largest scale size f...

Results obtained from experiments utilizing high‐power ion cyclotron range of frequencies (ICRF) heating to stabilize sawtooth oscillations on Tokamak Fusion Test Reactor (TFTR) [Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] are reviewed. The key observations include existence of a minimum ICRF power required to achieve stabilization, a dependence of the stabilization threshold on the relative size of the ICRF power deposition profile to the q=1 volume, and a peaking of the equilibrium pressure and current profiles during sawtooth‐free phases of the discharges. In addition, preliminary measurements of the poloidal magnetic field profile indicate that q on axis decreases to a value of 0.55±0.15 after a sawtooth‐stabilized period of ∼0.5 sec has transpired. The results are discussed in the context of theory, which suggests that the fast ions produced by the ICRF heating suppress sawteeth by stabilizing the m=1 magnetohydrodynamic (MHD) instabilities believed to be the trigger for the sawtooth oscillations. Though qualitative agreement is found between the observations and the theory, further refinement of the theory coupled with more accurate measurements of experimental profiles will be required in order to complete quantitative comparisons.

Measurements of the toroidal rotation speed vφ(r) driven by neutral beam injection in tokamak plasmas and, in particular, simultaneous profile measurements of vφ, Ti, Te, and ne, have provided new insights into the nature of anomalous transport in tokamaks. Low‐recycling plasmas heated with unidirectional neutral beam injection exhibit a strong correlation among the local diffusivities, χφ≊χi>χe. Recent measurements have confirmed similar behavior in broad‐density L‐mode plasmas. These results are consistent with the conjecture that electrostatic turbulence is the dominant transport mechanism in the tokamak fusion test reactor tokamak (TFTR) [Phys. Rev. Lett. 58, 1004 (1987)], and are inconsistent with predictions both from test‐particle models of strong magnetic turbulence and from ripple transport. Toroidal rotation speed measurements in peaked‐density TFTR ‘‘supershots’’ with partially unbalanced beam injection indicate that momentum transport decreases as the density profile becomes more peaked. In hi...

Magnetohydrodynamic (MHD) activity within three zones (core, half‐radius, and edge) of TFTR [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 51] tokamak plasmas are discussed. Near the core of the plasma column, sawteeth are often observed. Two types of sawteeth are studied in detail; one with complete, and the other with incomplete, magnetic reconnection. Their characteristics are determined by the shape of the q profile. Near the half‐radius the m/n=3/2 and 2/1 resistive ballooning modes are found to correlate with a beta collapse. The pressure and the pressure gradient at the mode rational surface are found to play an important role in stability. MHD activity is also studied at the plasma edge during limiter H modes. The edge localized modes (ELM’s) are found to have a precursor mode with a frequency between 50–200 kHz and a mode number m/n=1/0. The mode does not show a ballooning structure. While these instabilities have been studied on many other machines, ...

Co‐ and counter‐viewing bolometers aimed along a common tangency chord are being used to study power losses due to charge exchange (CX) of fast ions in neutral beam injection (NBI) heated TFTR plasmas. For unidirectional injection, tangential bolometers oriented to view CX loss of circulating fast ions detect losses from the thermal target plasma (impurity radiation and CX) plus power due to the fast ion CX loss, whereas bolometers oppositely directed measure only the target plasma contribution. The difference between the two signals is a measure of the fast ion CX loss. Additional information is obtained by comparing the tangential bolometer signals with those of perpendicularly viewing bolometer monitors and arrays. The measurements are compared to results of the TRANSP code analysis.

Plasmas have been formed in the tokamak fusion test reactor (TFTR) which separate completely from the limiter and are called detached plasmas. The physical evidence for the separation from the limiter is that all the ohmic input power is radiated away at the plasma boundary which can be 10→30 cm inside the limiter. The 10‐chord multichannel MIRI interferometer system has been used to diagnose this region and follows the formation of the detached state through a MARFE. MIRI indicates that a plasma ∼0.5→1×1013 cm−3 occurs in the region between the luminous boundary and the limiter. This work was supported by U.S. DOE Contract No. DE‐AC02‐76‐CHO‐3073.