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A study on the neutron production from single and double gas puff Z-pinches on the CESZAR linear transformer driver with ∼0.45 MA current and 170 ns rise time is presented. Total neutron yield measurements made with a LaBr activation detector are compared for three configurations, using a double nozzle setup. When a single, hollow, deuterium gas shell was used, reliable implosions could only be attained at higher load mass than the optimal value to match implosion time with the driver rise time, with neutron yields of ∼106 per pulse. The use of a double gas puff configuration with a deuterium center jet allowed a reduction in the shell density and operation closer to machine-matched conditions, recording up to (4.1 ± 0.3) × 107 neutrons/pulse when either Kr or D2 was used in the shell. For a comparable mass and implosion time, using a higher atomic-number gas in the outer shell results in more unstable plasma surface and smaller plasma radius at the location of instability bubbles, which, however, do not seem to consistently correlate with a higher neutron yield. Comparing implosion dynamics with models and neutron yields with literature scaling suggests that the machine current is not well coupled to the plasma during the final stages of compression. Optimizing current and energy coupling to the pinched plasma is critical to improving performance, particularly in low-impedance drivers. [ABSTRACT FROM AUTHOR]

An economic magnetic fusion reactor favours a high ratio of plasma kinetic pressure to magnetic pressure in a well-confined, hot plasma with low thermal losses across the confining magnetic field. Field-reversed configuration (FRC) plasmas are potentially attractive as a reactor concept, achieving high plasma pressure in a simple axisymmetric geometry. Here, we show that FRC plasmas have unique, beneficial microstability properties that differ from typical regimes in toroidal confinement devices. Ion-scale fluctuations are found to be absent or strongly suppressed in the plasma core, mainly due to the large FRC ion orbits, resulting in near-classical thermal ion confinement. In the surrounding boundary layer plasma, ion- and electron-scale turbulence is observed once a critical pressure gradient is exceeded. The critical gradient increases in the presence of sheared plasma flow induced via electrostatic biasing, opening the prospect of active boundary and transport control in view of reactor requirements.

Fast-ion D(α) measurements are an application of visible charge-exchange recombination (CER) spectroscopy that provide information about the energetic ion population. Like other CER diagnostics, the standard intensity calibration is obtained with an integrating sphere during a vacuum vessel opening. An alternative approach is to create plasmas where the fast-ion population is known, then calculate the expected signals with a synthetic diagnostic code. The two methods sometimes agree well but are discrepant in other cases. Different background subtraction techniques and simultaneous measurements of visible bremsstrahlung and of beam emission provide useful checks on the calibrations and calculations.

A neutral particle bolometer (NPB) has been designed and implemented on Tri Alpha Energy's C-2 device in order to spatially and temporally resolve the charge-exchange losses of fast-ion populations originating from neutral beam injection into field-reversed configuration plasmas. This instrument employs a silicon photodiode as the detection device with an integrated tungsten filter coating to reduce sensitivity to light radiation. Here we discuss the technical aspects and calibration of the NPB, and report typical NPB measurement results of wall recycling effects on fast-ion losses.

A code thatmodels signals produced by charge-exchange reactions between fast ions and injected neutral beams in tokamak plasmas is described. With the fastion distribution function as input, the code predicts the efflux to a neutral particle analyzer (NPA) diagnostic and the photon radiance of Balmer-alpha light to a fastion Dá (FIDA) diagnostic. Reactions with both the primary injected neutrals and with the cloud of secondary "halo" neutrals that surround the beam are treated. Accurate calculation of the fraction of neutrals that occupy excited atomic states (the collisionalradiative transition equations) is an important element of the code. Comparison with TRANSP output and other tests verify the solutions. Judicious selection of grid size and other parameters facilitate efficient solutions. The output of the code has been validated by FIDA measurements on DIII-D but further tests are warranted. © 2011 Global-Science Press.

A hot stable field-reversed configuration (FRC) has been produced in the C-2 experiment by colliding and merging two high-beta plasmoids preformed by the dynamic version of field-reversed theta-pinch technology. The merging process exhibits the highest poloidal flux amplification obtained in a magnetic confinement system (over tenfold increase). Most of the kinetic energy is converted into thermal energy with total temperature (T(i) + T(e)) exceeding 0.5 keV. The final FRC state exhibits a record FRC lifetime with flux confinement approaching classical values. These findings should have significant implications for fusion research and the physics of magnetic reconnection.

Combined neutral beam injection and high-harmonic fast-wave (HHFW) heating accelerate deuterium fast ions in the National Spherical Torus Experiment (NSTX). With 1.1 MW of HHFW power, the neutron emission rate is about three times larger than in the comparison discharge without HHFW heating. Acceleration of fast ions above the beam injection energy is evident on an E||B type neutral particle analyzer (NPA), a 4-chord solid state neutral particle analyzer (SSNPA) array and a 16-channel fast-ion D-alpha (FIDA) diagnostic. The accelerated fast ions observed by the NPA and SSNPA diagnostics mainly come from passive charge exchange reactions at the edge due to the NPA/SSNPA localization in phase space. The spatial profile of accelerated fast ions that is measured by the FIDA diagnostic is much broader than in conventional tokamaks because of the multiple resonance layers and large orbits in NSTX. The fast-ion distribution function calculated by the CQL3D Fokker-Planck code differs from the measured spatial profile, presumably because the current version of CQL3D uses a zero-banana-width model. In addition, compressional Alfven eigenmode activity is stronger during the HHFW heating and it may affect the fast-ion spatial profile. © 2010 IOP Publishing Ltd.

Combined neutral beam injection and fast wave heating at cyclotron harmonics accelerate deuterium fast ions in the National Spherical Torus Experiment (NSTX) and in the DIII-D tokamak. Acceleration above the injected energy is evident in fast-ion D-alpha (FIDA) and volume-average neutron data. The FIDA diagnostic measures spatial profiles of the accelerated fast ions. In DIII-D, the acceleration is at a 4th or 5th cyclotron harmonic; the maximum enhancement in the high-energy FIDA signal is 8-10 cm beyond the resonance layer. In NSTX, acceleration is observed at five harmonics (7-11) simultaneously; overall, the profile of accelerated fast ions is much broader than in DIII-D. The energy distribution predicted by the CQL3D Fokker-Planck code agrees fairly well with measurements in DIII-D. However, the predicted profiles differ from experiment, presumably because the current version of CQL3D uses a zero-banana-width model. © 2009 American Institute of Physics.

Neutral beam injection into a plasma with negative central shear produces a rich spectrum of toroidicity-induced and reversed-shear Alfvén eigenmodes in the DIII-D tokamak. The application of fast-ion D α (FIDA) spectroscopy shows that the central fast-ion profile is flattened in the inner half of the discharge. Neutron and equilibrium measurements corroborate the FIDA data. The temporal evolution of the current profile is also strongly modified. Studies in similar discharges show that flattening of the profile correlates with the mode amplitude and that both types of Alfvén modes correlate with fast-ion transport. Calculations by the ORBIT code do not explain the observed fast-ion transport for the measured mode amplitudes, however. Possible explanations for the discrepancy are considered. © 2008 IAEA, Vienna.

The Dα light emitted by neutralized deuterium fast ions is measured in magnetohydrodynamics (MHD)-quiescent, magnetically confined plasmas during neutral beam injection. A weighted Monte Carlo simulation code models the fast-ion Dα spectra based on the fast-ion distribution function calculated classically by TRANSP [R. V. Budny, Nucl. Fusion 34, 1247 (1994)]. The spectral shape is in excellent agreement and the magnitude also has reasonable agreement. The fast-ion Dα signal has the expected dependencies on various parameters including injection energy, injection angle, viewing angle, beam power, electron temperature, and electron density. The neutral particle diagnostic and measured neutron rate corroborate the fast-ion Dα measurements. The relative spatial profile agrees with TRANSP and is corroborated by the fast-ion pressure profile inferred from the equilibrium. © 2007 American Institute of Physics.

The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving βT and βN values of 17% and 5.7 (%m T MA-1), respectively. A newly available motional Stark effect diagnostic has enabled validation of current-drive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n ≤ 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable β in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection. © 2007 IAEA.

Reversed shear Alfv́n eigenmodes (RSAEs) are typically thought of as being localized near the minima in the magnetic safety factor profile, however, their spatial coupling to global toroidal Alfv́n eigenmodes (TAEs) has been observed in DIII-D discharges. For a decreasing minimum magnetic safety factor, the RSAE frequency chirps up through that of stable and unstable TAEs. Coupling creates a small gap at the frequency degeneracy point forming two distinct global modes. The core-localized RSAE mode structure changes and becomes temporarily global. Similarly, near the mode frequency crossing point, the global TAE extends deeper into the plasma core. The frequency splitting and spatial structure of the two modes throughout the various coupling stages, as measured by an array of internal fluctuation diagnostics, are in close agreement with linear ideal MHD calculations using the NOVA code. The implications of this coupling for eigenmode stability is also investigated and marked changes are noted throughout the coupling process. © 2007 American Institute of Physics.

We report on one-dimensional simulations of a staged Z-pinch (SZP) fusion scheme using the radiation-hydrodynamics code HYDRA [Marinak et al., Phys. Plasmas 8, 2275 (2001)]. A pulsed-power driver, based on the Z-machine at Sandia National Laboratories, with a peak current around 20 MA, is used to implode a high-Z gas liner surrounding a cylindrical DT fuel target. For the best conditions used in this work, excluding some physical processes, such as mix, we find a "clean" fusion yield of 7.05 MJ (2.5 × 10 18 neutrons), thus showing the potential for high yields using this scheme. The design achieving this yield was determined through a study of different liner materials, liner masses, and fuel masses. We also address previous simulations of the SZP setup that used ad hoc limits of the total internal energy of the liner that can artificially increase fusion yields. [ABSTRACT FROM AUTHOR]

The fast-ion distribution function in the National Spherical Torus Experiment is modified from shot to shot while keeping the total injected power at ∼2 MW. Deuterium beams of different energy and tangency radius are injected into helium L-mode plasmas, producing a rich set of instabilities, including compressional Alfvén eigenmodes, toroidicity-induced Alfvén eigenmodes (TAE), 50-100 kHz instabilities with rapid frequency sweeps or chirps, and strong, low frequency (10-20 kHz) fishbones. The experiment was motivated by a theory that attributes frequency chirping to the formation of holes and clumps in phase-space. In the theory, increasing the effective collision frequency of the fast ions that drive the instability can suppress frequency chirping. In the experiment, high-power (≲3 MW) high harmonic fast wave (HHFW) heating accelerates the fast ions in an attempt to alter the nonlinear dynamics. Steady-frequency TAE modes diminish during the HHFW heating but there is little evidence that frequency chirping is suppressed. © 2006 IOP Publishing Ltd.

The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving Βt ∼40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ∼2.8 and triangularity δ∼0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S≡ q95 Ip (a Bt), which has been observed at large values of the S∼37 [MA (m·T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip =1 MA, and for 1.6 s for Ip =700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion. © 2006 American Institute of Physics.

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6 - approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βt with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented. © 2006 IAEA, Vienna.

Alfvén instabilities excited by energetic ions are used as a means to reduce the central magnetic shear in a tokamak via redistribution of energetic ions. When the central magnetic shear is low enough, ballooning modes become stable for any plasma pressure gradient and an internal transport barrier (ITB) with a steep pressure gradient can exist. This mechanism can sustain a steady-state ITB as demonstrated by experimental data from the DIII-D tokamak. It can also produce a shear in toroidal and poloidal plasma rotation. Possible application of this technique to use the energetic α particles for improvement of burning plasma performance is discussed.

Hydrogenic fast-ion populations are common in toroidal magnetic fusion devices, especially in devices with neutral beam injection. As the fast ions orbit around the device and pass through a neutral beam, some fast ions neutralize and emit Balmer-alpha light. The intensity of this emission is weak compared with the signals from the injected neutrals, the warm (halo) neutrals and the cold edge neutrals, but, for a favourable viewing geometry, the emission is Doppler shifted away from these bright interfering signals. Signals from fast ions are detected in the DIII-D tokamak. When the electron density exceeds ̃7 × 1019 m-3, visible bremsstrahlung obscures the fast-ion signal. The intrinsic spatial resolution of the diagnostic is ̃5cm for 40keV amu-1 fast ions. The technique is well suited for diagnosis of fast-ion populations in devices with fast-ion energies (̃30 keV amu-1), minor radii (̃0.6 m) and plasma densities (≲ 1020 m-3) that are similar to those of DIII-D.

Alfvén instabilities can reduce the central magnetic shear via redistribution of energetic ions. They can sustain a steady state internal transport barrier as demonstrated in this DIII-D tokamak experiment. Improvement in burning plasma performance based on this mechanism is discussed.

An instability with a lower frequency than the toroidicity-induced Alfvén eigenmode was initially identified as a beta-induced Alfvén eigenmode (BAE). Instabilities with the characteristic spectral features of this "BAE" are observed in a wide variety of tokamak plasmas, including plasmas with negative magnetic shear. These modes are destabilized by circulating beam ions and they transport circulating beam ions from the plasma core. The frequency scalings of these "BAEs" are compared to theoretical predictions for Alfvén modes, kinetic ballooning modes, ion thermal velocity modes, and energetic particle modes. None of these simple theories match the data. © 1999 American Institute of Physics.

The Tokamak Fusion Test Reactor (TFTR) is a large tokamak which has performed experiments with 50:50 deuterium-tritium fuelled plasmas. Since 1993, TFTR has produced about 1090 D-T plasmas using about 100 grams of tritium and producing about 1.6 GJ of D-T fusion energy. These plasmas have significant populations of 3.5 MeV alphas (the charged D-T fusion product). TFTR research has focused on alpha particle confinement, alpha driven modes, and alpha heating studies. Maximum D-T fusion power production has aided these studies, requiring simultaneously operation at high input heating power and large energy confinement time (to produce the highest temperature and density), while maintaining low impurity content. The principal limitation to the TFTR fusion power production was the disruptive stability limit. Secondary limitations were the confinement time, and limiter power handling capability. © 1997 IOP Publishing Ltd.

Experiments in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 2, 2176 (1995)] have explored several novel regimes of improved tokamak confinement in deuterium–tritium (D–T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high [formula omitted]). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA [formula omitted] accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-[formula omitted] plasmas produced by rapid expansion of the minor cross section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D–T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D–T plasmas with [formula omitted] and weak magnetic shear in the central region, a toroidal Alfvén eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions. © 1997, American Institute of Physics. All rights reserved.

A search for alpha driven beta-induced Alfvén eigenmodes (BAEs) was conducted in low current (1.0-1.6 MA) TFTR supershots. Stable high beta deuterium-tritium (DT) discharges were obtained with βp = 2.4 and a central alpha beta of 0.1%. Instabilities between 75 and 200 kHz were observed by magnetics probes in many DT discharges, but the activity was also present in deuterium-deuterium (DD) comparison discharges, indicating that these modes are not destabilized (principally) by the alpha particle population. Losses of fusion products are also similar in the two sets of discharges. Theoretical simulations confirm that the achieved alpha particle pressure is too small to produce instability.

Because alpha particle losses can have a significant influence on tokamak reactor viability, the loss of deuterium–tritium alpha particles from the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] has been measured under a wide range of conditions. In TFTR, first orbit loss and stochastic toroidal field ripple diffusion are always present. Other losses can arise due to magnetohydrodynamic instabilities or due to waves in the ion cyclotron range of frequencies. No alpha particle losses have yet been seen due to collective instabilities driven by alphas. Ion Bernstein waves can drive large losses of fast ions from TFTR, and details of those losses support one element of the alpha energy channeling scenario. © 1996, American Institute of Physics. All rights reserved.

We present evidence that the fast-ion diffusion coefficient changes across the TFTR plasma column. Two MHD quiescent discharges are analyzed: a high power D-T plasma heated with 21 MW of deuterium and tritium beams, and an ohmic plasma into which a 10 MW, 20 ms deuterium beam pulse was injected. The localized charge-exchange measurements in the ohmic plasma, and the neutron flux measurements in the D-T plasma are compared with predictions from the 1 1/2 D transport simulation code TRANSP. We have modified the code to allow modelling with arbitrary fast-ion diffusion profiles Df(r). Significant improvement in the agreement between measurement and simulation is obtained with a Df(r) profile that has low values in the inner half of the plasma column (Df

A search is made for signs of enhanced beam ion transport in a set of TFTR plasmas heated with various numbers of tritium and deuterium beams. Our investigation is based on the time dependent, 11/2-D simulation code TRANSP, and an ad hoc diffusion model of enhanced fast ion transport. Spatially constant and spatially variable diffusion models are used (the code was upgraded to allow for Df(r) modelling). Simulations with spatially constant diffusion coefficients resulted in an upper bound of Df 0.2 m2/s for a set of high power DT supershots. One discharge is analysed in particular detail, in terms of both various diffusion models and the effects of systematic errors. Best agreement between the measured neutron flux in this discharge and the corresponding TRANSP predictions is obtained by assuming that Df has low values in the inner half of the plasma column (Df 0.1 m 2/s) and then rises rapidly in the outer half, suggesting that stochastic ripple diffusion is the likely mechanism for the enhanced beam ion transport. This hypothesis is supported by the modelling results from two other plasmas with large major radii.

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date (September 1995), the maximum fusion power has been 10.7 MW, using 39.5 MW of neutral beam heating, in a supershot discharge and 6.7 MW in a high beta P discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MW/m3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER). The energy confinement time tau E is observed to increase in DT, relative to D plasmas, by 20% and the n1(0).T1(0). tau E 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 beta P discharges. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations assuming classical confinement. Measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from helium gas puffing experiments. The loss of energetic 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. ICRF heating of a DT plasma, using the second harmonic of tritium, has been demonstrated. DT experiments on TFTR will continue both to explore the physics underlying the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit.

A detector system based on partially depleted silicon surface barrier detectors and fast front-end electronics has been built and cross calibrated to a set of absolutely calibrated 4He recoil detectors. The cross-calibration factor for the channel with the widest dynamic range is 2.5×10-13 counts per 14 MeV source neutron. These data agree well with the independent neutron activation data. The new detector system covers a large dynamic range (corresponding to 1013-1018 neutrons/s). The response is linear, except at the highest count rates where the detector dead time (∼200 ns) causes departure from linearity. The noise discrimination against 2.5 MeV neutrons and γ pileup is excellent. Measurements of D-T neutrons from a tritium gas puff experiment as well as from a high-power D-T discharge in the TFTR tokamak are presented. © 1995 American Institute of Physics.

Various DD and DT plasmas are analysed for effects of fast ion transport with a time dependent, 1 1/2 -D transport simulation code (TRANSP). The sensitivity of the simulations to fast ion diffusion modelling is tested against numerous parameters. Strong correlations are found with beam power and plasma stored energy. The neutron emission sensitivity is mostly affected by the fraction of beam-beam neutrons. Wall recycling is essential in interpreting the results for DT plasmas heated with pure deuterium or pure tritium beams. The decay of the 14 MeV neutron emission following a short DT beam pulse implies a small fast ion diffusion coefficient (Df

During the initial deuterium-tritium experiments on TFTR, neutron emission was measured with 235U and 238U fission chambers, silicon surface barrier diodes, spatially collimated 4He proportional counters and ZnS scintillators, and a variety of elemental activation foils. The activation foils, 4He counters, and silicon diodes can discriminate between 14 and 2.5 MeV neutrons. The other detectors respond to both DD and DT neutrons but are more sensitive to the latter. The proportional counters, scintillators, and some of the fission chambers were calibrated absolutely, using a 14 MeV neutron generator positioned at numerous locations inside the TFTR vacuum vessel. Although the directly calibrated systems were saturated during the highest-power deuterium-tritium operation, they allowed cross calibration of less sensitive fission chambers and silicon diodes. The estimated absolute accuracy of the uncertainty-weighted mean of these cross calibrations, combined with an independent calibration derived from activation foil determinations of total neutron yield, is ±7%. © 1995 American Institute of Physics.

Three campaigns, prior to July 1994, attempted to increase the fusion power in DT plasmas on the Tokamak Fusion Test Reactor (TFTR). The first campaign was dedicated to obtaining >5 MW of fusion power while avoiding MHD events similar to the JET X-event. The second was aimed at producing maximum fusion power irrespective of proximity to MHD limits, and achieved 9 MW limited by a disruption. The third campaign increased the energy confinement time using lithium pellet conditioning while raising the ratio of alpha heating to beam heating.