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The paper deals with time domain-deterministic stochastic assessment of a transient electric field generated by a ground penetrating radar (GPR) dipole antenna and transmitted into a lower half-space. The deterministic time domain formulation is based on the space-time Hallen integral equation for half-space problems. The Hallen equation is solved via the Galerkin–Bubnov variant of the Indirect Boundary Element Method (GB-IBEM) and the space-time current distribution along the dipole antenna is obtained, thus providing the field calculation. The field transmitted into the lower medium is obtained by solving the corresponding field integrals. As GPR systems are subjected to a rather complex environment, some input parameters, for example the antenna height over ground or earth properties, are partly or entirely unknown and, therefore, a simple stochastic collocation (SC) method is used to properly access relevant statistics about GPR time responses. The SC approach also aids in the assessment of corresponding confidence intervals from the set of obtained numerical results. The expansion of statistical output in terms of mean and variance over a polynomial basis, via the SC method, is shown to be a robust and efficient approach providing a satisfactory convergence rate.

A straight wire antenna model of the vertical grounding electrode has been presented in the paper. The formulation is based on the homogeneous integral equation of Pocklington type for half-space problems. The influence of a finitely conducting half-space is taken into account via the corresponding reflection coefficients. The Pocklington equation is solved both numerically and analytically. The numerical solution is carried out via the Galerkin-Bubnov scheme of the Indirect Boundary Element Method (GB-IBEM). Some illustrative computational examples are presented at the end.

The physics of current profile modification with lower hybrid waves has been investigated in the Princeton Beta Experiment-Modification tokamak by comparing experimental internal magnetic field profiles with ray tracing code predictions. When the n| spectrum of the launched waves was varied, local changes in the current profile were observed according to equilibria, reconstructed from motional Stark effect polarimetry measurements. Changes in the central safety factor (q) were also determined to be a function of the applied radiofrequency (RF) power. Calculations of the non-inductive current with the Tokamak Simulation Code/Lower Hybrid Simulation Code were unable to duplicate the measured current profiles. Instead, a current broadening mechanism like a finite fast electron diffusion coefficient had to be hypothesized to fit the data. This study was possible only with recent improvements in diagnostics and numerical analysis tools, and it constitutes an important contribution to our understanding of non-inductive current drive for future fusion reactors

A database for the 1993 run period of PBX-M has been documented (i), to investigate whether it was possible to forecast the development of MHD instabilities often observed in LH-assisted discharges and (ii), to detail the origin, the nature and the effects of those instabilities. The deposition radius of the RF current, the plasma internal inductance and the LH power are used to separate MHD active and quiescent regimes prior to the MHD onset. 1/1, 2/1, 3/1 global modes driven by the m = 2, n = 1 component are observed in discharges with LHCD. The destabilization is attributed to an increase of the current density gradient within the q = 2 surface. MHD fluctuations reduce the soft X-ray and hard X-ray intensities mainly around the RF current deposition radius. Minor disruptions with large inversion radius and mode locking are analysed. A possible precursor to the MHD is evidenced on the hard X-ray horizontal profiles. A resonance between fast trapped electrons and turbulent waves present in the background plasma is proposed to support the observations.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first‐orbit loss mechanism. No loss due to alpha‐particle‐driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

In many of the medium and high beta discharges in PBX-M, low n modes (n=1-4) with different n numbers are observed. The prevalence of low n modes decreases with increasing n number. Furthermore, if two modes of different frequency and n number (?1, and ?2; k1 and k2) are simultaneously present in the plasma, these modes interact non-linearly and create sidebands in frequency (?2 ? ?1) and wavenumber (k2 ? k1 or n2 ? n1 and m2 ? m1). It has also been observed that if these fundamental modes, ?1/k1 and ?2/k2, contain strong harmonics, the harmonics also interact non-linearly, creating more non-linear products: k?2 ? l?1 and kk2 ? lk1, where k and l are integers describing the harmonics. These multiple three wave couplings on a given rational surface can create several modes of the same m and n numbers but of different phase velocities. The multiple perturbations of the rational surfaces can increase the plasma losses. In fact, a decrease in neutron rate and an enhanced loss of medium energy ions are observed during this three wave coupling interaction. Although the three wave coupling has been observed primarily in medium or high beta discharges, it is a more general plasma phenomenon because it also occurs occasionally at low ?

In PBX-M, periods of enhanced transport are often observed in high beta polH-mode discharges. They can reduce the energy confinement time by a factor of two. Their duration is typically between 2 to 20 ms. They are followed by periods of normal transport of a similar duration. These periods of enhanced transport consist of a series of turbulent m=even MHD events, starting several cm inside the plasma edge. They propagate into the confinement region in 600 mu s, causing a progressive erosion of the profile. Each m=even event can also cause an energy loss of 1% or less, mainly from the confinement region. These m=even events are reminiscent of grassy ELMs.

Edge Localized Modes, or ELMs, are at the same time both disadvantageous and advantageous features of H-modes. ELMs can cause energy and particle losses from the plasma, with instantaneous energy losses reaching 40% of the total stored energy in the case of large-amplitude ELMs. The energy loss from these ELMs, or closely spaced smaller ELMs, may result in what is known as a beta-collapse, an event in which energy is lost over a substantial duration of the discharge. On the other hand, if controlled, the ELM related energy and, especially, impurity losses can lead to near steady-state discharge conditions with relatively low radiative power losses. The importance of ELMs in the development of the H-phase has led to numerous studies which attempt to uncover their nature. The purpose of the present work is to present additional observations of, most notably, high frequency magnetic fluctuations associated with ELMS. There are some observations here that are consistent with previously reported ones, namely the appearance of high frequency ({ge}250 kHz) activity prior to the D{sub {alpha}} increase, suggesting again that the rise in D{sub {alpha}} is some delayed response to an instability occurring farther inside the plasma edge. While this high frequency activitymore » is global, with high coherence for all coil pairs both on the inside and outside midplane, there is no identifiable mode structure. Reported here for the first time are observations of a very intense outward ballooning feature that appears simultaneously with the D{sub {alpha}} increase and lasts for about 5 msec. The toroidal mode structure of this feature is clearly identifiable, consisting of a series of toroidal harmonics from n=3 to 8, at frequencies given by f{sub n}=f{sub 0}+n{Delta}f, with f{sub 0}=30 to 50 kHz and A{Delta}=20 to 30 kHz.« less

The PBX-M tokamak has an array of unique features that allow it to study regimes of enhanced confinement and stability. In particular, because of its divertor operation, PBX-M has proved to be a valuable contributor to identifying and understanding various features of H-mode physics. To be reported here are recent results of H-mode studies of m = 1 bias induced H-modes, L- to H-transitions, including suppression of turbulent transport, ELMs and other related energy loss mechanisms, and Ion Bernstein Wave modification of H-mode discharges

The experimental program of the Princeton Beta Experiment‐Modified (PBX‐M) tokamak [Phys. Fluids B2, 1271 (1990)] is directed toward tailoring plasma profiles to achieve greater stability and confinement and to gain access to the second stability region. Modification of the current density profile has been achieved with lower‐hybrid current drive (LHCD), leading to a regime free of global magnetohydrodynamic modes, while raising the value of q(0) above unity. The diffusion of the fast electrons produced by LHCD has been examined using two‐dimensional hard x‐ray imaging. Ion Bernstein waves (IBW) have been used for ion heating: a preliminary analysis shows that ion heating was spatially localized and in agreement with theoretical calculations. Divertor biasing has modified the electric field inside the last closed surface, resulting in the formation of a transport barrier, which in turn has reduced the threshold power of neutral beam injection (NBI) for H‐mode transition by 25%.

The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m=1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions.

Through semi-independent absolute calibrations of multiply redundant neutron detector systems, the Tokamak Fusion Test Reactor (TFIR) has achieved {+-}7% (one-sigma) accuracy in its fusion power measurements. This has required careful attention to the linearity of detectors up to the present highest fusion power levels achieved on TFTR of over 10 MW. The extended duration of the DT program on TFTR has also tested the stability of the detector systems. These issues of calibration, linearity, and stability are reviewed for the TFTR experience and how it can be applied to plans for ITER.

Lower hybrid current drive (LHCD) is being explored as a means to control the current density profile on PBX‐M with the goal of raising the central safety factor q(0) to values of 1.5–2 to facilitate access to a full‐volume second stable regime. Initial experiments have been conducted with up to 400 kW of 4.6 GHz LH power in circular and indented plasmas with modest parameters. A tangential‐viewing two‐dimensional hard x‐ray imaging diagnostic has been used to observe the bremsstrahlung emission from the suprathermal electrons generated during LHCD. Hollow hard x‐ray images have indicated off‐axis localization of the driven current. A serious obstacle to the control of the current density profile with LHCD is the concomitant generation of MHD activity, which can seriously degrade the confinement of suprathermal electrons. By combining neutral beam injection with LHCD, an MHD‐free condition has been obtained where q(0) is raised above 1.

Understanding the wave damping mechanism in the presence of a `spectral gap` is an important issue for the current profile control using Lower Hybrid Current Drive (LHCD). The authors examine a traditional explanation based upon upshifting of the wave parallel refractive index (n{sub {parallel}}) and find that there can be an upper bound in the n{sub {parallel}} upshift. The amount of upshift is not sufficient to bridge the spectral gap completely under some PBX-M LHCD conditions. There is experimental evidence, however, that current was driven even under such conditions. Another mechanism is also considered, based upon the 2-D velocity space dynamics coupled with a compound wave spectrum, here consisting of forward- and backward-running waves. The runaway critical speed relative to the phase speeds of these waves plays an important role in this model.

Plasmas exhibiting the ``fishbone`` instability studied on the PBX-M tokamak show a distinct relationship between the plasma shape, the internal magnetic structure, and the presence or absence of fast ion losses associated with the fishbone mode. We have, for the first time, carried out measurements of the magnetic safety factor profile in fishbone-unstable plasmas, and used the knowledge of the associated experimental equilibria to compare the stability and fast ion loss properties of these plasmas with experimental observations.

High {beta}{sub pol} H-mode plasmas in the PBX-M tokamak often exhibit periods of Edge Localized Mode (ELM) activity, with each ELM preceded by a short duration ({le} 350 {mu}sec) burst of high frequency (200 to 250 kHz) magnetic activity. The burst grows on a time scale of 10 {mu}sec, and disappears just prior to the rapid increase in the D{sub {alpha}} emission that is characteristic of the ELM. The burst of activity is observed at all poloidal locations, with the largest amplitudes seen on the coils on the inner major radius side, indicating that the mode is not outward ballooning in character. Stability calculations indicate that a likely candidate for this high frequency ELM precursor is the pressure-driven ideal kink. 12 refs., 4 figs.

A particle scoop limiter with a graphite face backed by a 50 liter volume for collecting particles was used in PDX. Experiments were performed to test its particle control and power handling capabilities with up to 5 MW of D° power injected into D+ plasmas. Line average plasma densities of up to 8 × 1013 cm−3 and currents up to 450 kA were obtained. Plasma densities in the scoop channels greater than 2 × 1013 cm−3 and neutral densities in the scoop volume greater than 5 × 1014 cm−3 were observed. There is evidence that recycling may have occurred in the scoop channels for several discharges with large line-averaged plasma density. At beam powers up to 2.5 MW, energy confinement times above 40 ms were deduced from magnetics measurements and from transport analysis. Pressures in the vacuum vessel were in the 10 −5 Torr range, and recycling source neutral densities in the central plasma were low.

Impurity transport has been studied in ohmic- and neutral-beam-heated TFTR discharges by observation and numerical modeling of time evolutions of emissions from germanium injected using the laser-blowoff method. VUV spectral lines, the total radiated power, and soft x-ray intensities are compared with code predictions in which the impurity flux is assumed to have diffusive and convective terms. In plasmas with I/sub p/=1.4 MA and anti n/sub e/approx. = 2.5 x 10/sup 13/ cm/sup -3/, little difference in transport between ohmic and beam-heated discharges is observed. At I/sub p/=0.7 MA and anti n/sub e/approx. = 1.0 x 10/sup 13/ cm/sup -3/, a three-fold increase in the diffusion coefficient is found during beam heating.

Neutral-beam heating of plasmas in the Tokamak Fusion Test Reactor at low preinjection densities (n/sub e/(0)approx. =10/sup 19/ m/sup -3/) were characterized by T/sub e/(0) = 6.5 keV, T/sub i/(0) = 20 keV, n/sub e/(0) = 7 x 10/sup 19/ m/sup -3/, tau/sub E/ = 170 msec, ..beta../sub t//sub h//sub e//sub t//sub a/ = 2, and a d(d,n)/sup 3/He neutron emission rate of 10/sup 16/ sec/sup -1/. The ion temperature and the deuterium-fusion neutron yields were significantly higher than for previous tokamak experiments. The low initial densities were achieved by operation of the Tokamak Fusion Test Reactor with low plasma currents (less than or equal to1 MA) and by extensive limiter conditioning.

A progress report on the experimental studies of vacuum vessel conditioning during the first year of TFTR operation is presented. A previous paper described the efforts expended to condition the TFTR vessel prior to and during the initial plasma start-up experiments. During the start-up phase, discharge cleaning was performed with the vessel at room temperature. For the second phase of TFTR operations, which was directed towards the optimization of ohmically-heated plasmas, the vacuum vessel could be heated to 150°C. The internal configuration of the TFTR vessel was more complex during the second phase with the addition of a TiC/C moveable limiter array, inconel bellows cover plates, and ZrAl getter pumps. A quantitative comparison is given on the effectiveness of vessel bakeout, glow discharge cleaning, and pulse discharge cleaning in terms of total quantity of removed carbon and oxygen, residual gas base pressures and the resulting plasma impurity levels as measured by visible, UV and soft X-ray spectroscopy. The initial experience with hydrogen isotope changeover in TFTR is presented including the results of the attempt to hasten the changeover-time by using a glow discharge to precondition the vessel with the new isotope.

Impurity studies in L- and H-mode discharges in PBX have shown that both types of discharges can evolve into either an impurity accumulative or nonaccumulative case. In a typical accumulative discharge, Z eff peaks in the center to values of about 5. The central metallic densities can be high, n met / n e ≈ 0.01, resulting in central radiated power densities in excess of 1 W/cm 3 , consistent with bolometric estimates. The radial profiles of metals obtained independently from the line radiation in the soft X-ray and the VUV regions are very peaked. Concurrent with the peaking, an increase in the impurity influx coming from the edge of the plasma is observed. At the beginning of the accumulation phase the inward particle flux for titanium has values of 6 × 10 10 and 10 × 10 10 particles / cm 2 s at minor radii of 6 and 17 cm. At the end of the accumulation phase, this particle flux is strongly increased to values of 3 × 10 12 and 1 × 10 12 particles / cm 2 s. This increased flux is mainly due to influx from the edge of the plasma and to a lesser extent due to increased convective transport. Using the measured particle flux, an estimate of the diffusion coefficient D and the convective velocity ν is obtained.

A detector setup with three LN‐cooled Si(Li) diodes is used to measure soft x‐ray spectra (0.8–20 keV) emitted from the inside walls of the PDX vessel during the plasma discharge. The setup is part of a pulse‐height‐analysis system, which is used to measure the plasma and wall radiation simultaneously at five different radial positions. The wall and the plasma radiation are measured under different plasma conditions (e.g., OH and neutral beam heating). The wall radiation is very much increased during the neutral beam heating, with an enhancement factor of at least 10 over the OH wall radiation. Since we measure the plasma and the wall radiation at the same time, the measurements allow the conclusion that the wall radiation can be attributed essentially to fluorescence (line radiation, e.g., Ti‐Kα) and scattering (continuum part of the wall radiation spectrum). The fluorescence and the scattering are both caused by soft x‐ray radiation flux coming from the plasma. There seems to be no need to invoke other,...

The bremsstrahlung emission from the PLT tokamak during lower-hybrid current drive has been measured as a function of angle between the magnetic field and the emission direction. The emission is peaked strongly in the forward direction, indicating a strong anisotropy of the electron velocity distribution. The data demonstrate the existence of a nearly flat tail of the velocity distribution, which extends out to approximately 500 keV and which is interpreted as the plateau created by Landau damping of the lower-hybrid waves.

Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam -heated discharges. The TFTR tokamak has reached its original machine-design specifications ( I p = 2.5 MA and B T = 5.2 T). Initial neutral-beam -heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 10 20 m -3 in high-current discharges. At the opposite extreme, by operating at low plasm a current ( I p ~ 0.8 MA) and low density ( n e~ 1 x 10 19 m -3 ), high ion temperatures (9 + 2 keV) and rotation speeds (7 x 10 5 m s -1 ) have been achieved during injection. In addition, plasma-compression experiments have demonstrated acceleration of beam ions from 82 to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam -heated discharges. The result of these confinement studies are presented.

The PDX divertor configuration has recently been converted from an open to a closed geometry to inhibit the return of neutral gas from the divertor region to the main chamber. Since then, operation in a regime with high energy confinement in neutral beam heated discharges (ASDEX H-mode) has been routine over a wide range of operating conditions. These H-mode discharges are characterized by a sudden drop in divertor density and H α emission and a spontaneous rise in main chamber plasma density during neutral beam injection. The confinement time is found to scale nearly linearly with plasma current, but can be degraded due either to the presence of edge instabilities or heavy gas puffing. Detailed Thomson scattering temperature profiles show high values of T c near the plasma edge (∼ 450 eV) with sharp radial gradients (∼ 400 eV/cm) near the separatrix. Density profiles are broad and also exhibit steep gradients close to the separatrix.

This paper presents a summary of plasma-material interactions which influence the operation of TFTR with high current (≤ 2.2 MA) ohmically heated, and high-power (∼ 10 MW) neutral-beam heated plasmas. The conditioning procedures which are applied routinely to the first-wall hardware are reviewed. Fueling characteristics during gas, pellet, and neutral-beam fueling are described. Recycling coefficients near unity are observed for most gas fueled discharges. Gas fueled discharges after helium discharge conditioning of the toroidal bumper limiter, and discharges fueled by neutral beams and pellets, show R

The paper describes the present (end of February 1985) status of the plasma confinement studies in the TFTR tokamak with emphasis on those with neutral beam injection (NBI). Recent improvements in the device capabilities have substantially extended operating parameters: B /sub T/ increased to 4.0 T, I /sub p/ to 2.0 MA, injection power (P /sub b/ ) to /sup 5/ MW with H/sup 0/ or D/sup 0/ beams, n /sub e/ to 5 x 10/sup 19/ m/sup -3/ and Z /sub eff/ reduced to 1.4. With ohmic heating (OH) alone, the previously established scaling for gross energy confinement time (T /sub E/ /sup 0/ n /sub e/ q) has been confirmed at higher I /sub p/ and B /sub T/ , and the maximum T /sub E/ of 0.4 sec has been achieved. With NBI at P /sub b/ substantially (by factor > 2) higher than P /sub OH/ , excellent power and particle accountability have been established. This suggest that the less-than-expected increase in stored energy with NBI is not due to problems of power delivery, but due to problems of confinement deterioration. T /sub E/ is observed to scale approximately as I /sub p/ P /sub b/more » /sup 0.5/ (independent of n /sub e/ ), consistent with previous L-mode scalings. With NBI we have achieved the maximum T /sub E/ of 0.2 s and the maximum T /sub i/ (o) of 4.4 keV in the normal operating regime, and even higher T /sub i/ (o) in the energetic-ion regime with low-n /sub e/ and low I /sub p/ operation.« less

The PDX tokamak provides an experimental facility for the direct comparison of various impurity control techniques under reactor-like conditions. Four neutral beam lines inject > 6 MW for 300 ms. Carbon rail limiter discharges have been used to test the effectiveness of perpendicular injection, but non-disruptive full power operation for > 100 ms is difficult without extensive conditioning. Initial tests of a toroidal bumper limiter indicate reduced power loading and roughly similar impurity levels compared to the carbon rail limiter discharges. Poloidal divertor discharges with up to 5 MW of injected power are cleaner than similar circular discharges, and the power is deposited in a remote divertor chamber. High density divertor operation indicates a reduction of impurity flow velocity in the divertor and enhanced recycling in the divertor region during neutral injection.

Absorber foils have been installed in the TFTR X-Ray Imaging System to permit measurement of the electron temperature along 10 to 30 chords spaced at 5-12.5 cm with a time resolution of less than 100 ..mu..s. The technique uses the ratio of x-ray fluxes transmitted through two different foils. The ratio depends mainly on electron temperature. Simulations show that strong impurity line radiation can distort this ratio. To correct for these effects, special beryllium-scandium filters are employed to select the line-free region between 2 and 4.5 keV. Other filter pairs allow corrections for Fe L and Ni L line radiation as well as Ti K and Ni K emission. Good accuracy is also obtained with simple beryllium filters, provided that impurity corrections are incorporated in the analysis, taking line intensities from the x-ray pulse-height analysis diagnostic. A description of modeling calculations and a comparison of temperature values from this diagnostic with data from the x-ray pulse height analysis, the electron cyclotron emission, and the Thomson scattering diagnostics are presented. Several applications of the absorber foil electron temperature diagnostic on TFTR are discussed.

Charge-dependent convective impurity transport is observed in the central core of PBX H-mode discharges by measurement of radial profiles for both low- and high-Z intrinsic impurities in the presence of strong sawtooth activity. The time-evolving ${Z}_{\mathrm{eff}}$ profile and central metallic densities are modeled to yield a diffuison coefficient and a Z-dependent inward convective pinch which are in general agreement with estimates of neoclassical transport in the central plasma region.

Fueling requirements and impurity levels in neutral-beam-heated discharges in the PDX tokamak have been compared for plasmas formed with conventional graphite rail limiters, a particle scoop limiter, and an open or closed poloidal divertor. Gas flows necessary to obtain a given density are highest for diverted discharges and lowest for the scoop limiter. Hydrogen pellet injection provides an efficient alternative fueling technique, and a multiple pellet injector has produced high density discharges for an absorbed neutral beam power of up to 600 kW, above which higher speeds or more massive pellets are required for penetration to the plasma core. Power balance studies indicate that 30–40% of the total input power is radiated while ~15% is absorbed by the limiting surface, except in the open divertor case, where 60% flows to the neutralizer plate. In all operating configurations, Z eff usually rises at the onset of neutral beam injection. Both open divertor pl;asmas and those formed on a well conditioned water-cooled limiter have Z eff ⪅ 2 at the end of neutral injection. A definitive comparison of divertors and limiters for impurity control purposes requires longer beam pulses or higher power levels than available on present machines.

High resolution spectra of the Ly ..cap alpha../sub 1/ and Ly ..cap alpha../sub 2/ lines of hydrogenlike titanium, TiXXII, and the associated dielectronic satellites which are due to transitions 1snl-2pnl with n greater than or equal to 2, have been observed from tokamak discharges with auxiliary ion cyclotron heating (ICRH) with central electron temperatures of 2 keV and central electron densities of 8 x 10/sup 13/ cm/sup -3/ on the Princeton Large Torus (PLT). The data have been used for a detailed comparison with theoretical predictions based on the Z - expansion method and Hartree - Fock calculations. The results obtained with the Z - expansion method are in excellent agreement with the observed spectral data except for minor discrepancies between the theoretical and experimental wavelengths of 0.0003 A for the n = 2 satellites and of 0.0001 A for the separation of the Ly ..cap alpha../sub 1/ and Ly ..cap alpha../sub 2/ lines. Very good agreement with the experimental data is also obtained for the results from the Hartree - Fock calculations though somewhat larger discrepancies (approx. = 0.0009 A) exist between experimental and theoretical wavelengths which are systematically too small. The observed spectra are used for diagnosis ofmore » the central ion and electron temperatures of the PLT discharges and for a measurement of the dielectronic recombination rate coefficient of TiXXII.« less

TFTR pellet injection results reported by Schmidt (1985) have been extended to higher density and n tau in plasmas limited by a graphite inner-wall belt limiter. Increased pellet penetration and larger density increases were achieved by operation at reduced plasma current (1.6 MA) and minor and major radius (70 cm and 235 cm). Under these conditions, beam heating results have been extended to 7 MW.

The soft x‐ray continuum radiation in tokamak fusion test reactor (TFTR) low‐density neutral‐beam discharges can be much lower than its theoretical value obtained by assuming a corona equilibrium. This reduced continuum radiation is caused by an ionization equilibrium shift toward lower states, which strongly changes the value of the average recombination coefficient of metallic impurities γ, even for only slight changes in the average charge Z. The primary agent for this shift is the charge exchange between the highly ionized impurity ions and the neutral hydrogen, rather than impurity transport, because the central density of the neutral hydrogen is strongly enhanced at lower plasma densities with intense beam injection. In the extreme case of low‐density, high neutral‐beam power TFTR operation (energetic ion mode) the reduction in γ can be as much as one‐half to two‐thirds. We calculate the parametric dependence of γ and Z for Ti, Cr, Fe, and Ni impurities on neutral density (equivalent to beam power), electron temperature, and electron density. These values are obtained by using either a one‐dimensional impurity transport code or a zero‐dimensional code with a finite particle confinement time. As an example, we show the variation of γ and Z in different TFTR discharges.

Because soft‐x‐ray pulse‐height‐analysis spectra contain chordal information, the electron temperature and the radiation intensity (enhancement factor) measurements do not represent the local values. The correction factors for the electron temperature and the enhancement factor as a function of the temperature and density profile parameters and the energy are obtained. The spectrum distortion due to pulse pileup effects is also evaluated. A set of curves is given from which the distortion of the spectrum can be obtained if the electron temperature, the Be filter thickness, and the electronic parameters of the acquisition system are known.

Tangentially coinjected deuterium beam ions were accelerated from 82 up to 150 keV during a major-radius compression experiment in the tokamak fusion test reactor. The ion energy spectra and the variation in fusion yield were in good agreement with Fokker-Planck code simulations. In addition, the plasma rotation velocity was observed to rise during compression.