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The Princeton Plasma Physics Laboratory (PPPL), in collaboration with the Naval Research Laboratory (NRL), is currently investigating various novel materials (single crystal silicon, , and ) for use as electron-beam transmission windows in a KrF excimer laser system. The primary function of the window is to isolate the active medium (excimer gas) from the excitation mechanism (field-emission diodes). Chosen window geometry must accommodate electron energy transfer greater than 80% (750 keV), while maintaining structural integrity during mechanical load (1.3 to 2.0 atm base pressure differential, approximate 0.5 atm cyclic pressure amplitude, 5 Hz repetition rate) and thermal load across the entire hibachi area (approximate 0.9 W {center_dot} cm superscript ''-2''). In addition, the window must be chemically resistant to attack by fluorine free-radicals (hydrofluoric acid, secondary). In accordance with these structural, functional, and operational parameters, a 22.4 mm square silicon prototype window, coated with 500 nm thin-film silicon nitride (Si{sub 3}N{sub 4}), has been fabricated. The window consists of 81 square panes with a thickness of 0.019 mm {+-} 0.001 mm. Stiffened (orthogonal) sections are 0.065 mm in width and 0.500 mm thick (approximate). Appended drawing (Figure 1) depicts the window configuration. Assessment of silicon (and silicon nitride) materialmore » properties and CAD modeling and analysis of the window design suggest that silicon may be a viable solution to inherent parameters and constraints.« less

Experiments in the Tokamak Fusion Test Reactor (TFTR) 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-l{sub i}). 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 (q{sub a} {approx} 4.3), 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-l{sub i} 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 q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven 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.

Continued operation of the tokamak fusion test reactor (TFFR) with a mixture of deuterium and tritium fueling has permitted the opportunity to measure the retention of tritium in the graphite limiter and to investigate the use of discharge cleaning techniques and venting to remove the tritium. The tritium was introduced into TFTR by neutral beam injection and by gas puffing. The limiter is subject to erosion and codeposition. While short term retention was high, the retention averaged over the 1993–1995 D-T campaign was 52 ± 15%. The tritium removal techniques resulted in lowering the in-vessel inventory from 16.4 kCi (1 Ci = 2.076 × 1019 tritium atoms and 10 kCi = 1.04 g) at the end of 1995 operation to 7.2 kCi at the start of the 1996 experimental program.

The Tritium Purification System (TPS) is a hydrogen isotope separation system put into operation within the Tokamak Fusion Test Reactor (TFTR) Tritium Systems. The TPS operates in two stages: extraction of hydrogen isotopes from the TFTR plasma waste effluents via a Palladium/Silver diffuser; and separation of hydrogen isotopes via a multiple-stage cryogenic distillation system. Commissioning of TPS included: Operational testing at Canadian Fusion Fuels Technology Project (CFFTP) and at Princeton, thorough helium and tritium leakchecks, trial run with a limited tritium inventory (1 kCi), and an integrated systems test using 10 kCi of tritium. The integrated systems test, which was started in April of 1995 took approximately eight months to perform. Several `infant mortality` failures, requiring numerous line breaks into highly contaminated piping, were safely performed. On December 18, 1995 the TPS delivered its first batch of purified tritium product. This paper provides a brief overview of the TPS design and theory of operation. The focus of this paper is the commissioning, operation, performance and maintenance of the device. Lessons learned in maintenance and repair of the TPS are also addressed. 1 ref.

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.

TFTR has begun its campaign to study deuterium-tritium fusion under reactor-like conditions. Variable amounts of deuterium and tritium neutral beam power have been used to maximize fusion power, study alpha heating, investigate alpha particle confinement, and search for alpha driven plasma instabilities. Additional areas of study include energy and particle transport and confinement, ICRF heating schemes for DT plasmas, tritium retention, and fusion in high βp plasmas. The majority of this work is done in the TFTR supershot confinement regime. To obtain supershots, extensive limiter conditioning using helium fueled ohmic discharges and lithium pellet injection into ohmic and neutral beam heated plasmas is performed, resulting in a low recycling limiter. The relationship between recycling and core plasma confinement has been studied by using helium, deuterium and high-Z gas puffs to simulate high recycling limiter conditions. These studies show that confinement in TFTR supershots is very sensitive to the influx of neutral particles at the plasma edge.

The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107 kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9 MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments.

In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength Sn and D–D fusion power gain QDD are realized in the neutral‐beam‐fueled and heated ‘‘supershot’’ regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, Sn increases approximately as P1.8b. The highest‐Q shots are characterized by high Te (up to 12 keV), Ti (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad Te profiles, and lower Zeff. Replacement of critical areas of the graphite limiter tiles with carbon‐fiber composite tiles and improved alignment with the plasma have mitigated the ‘‘carbon bloom.’’ Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, QDD increases with decreasing pre‐injection carbon radiation, and increases strongly with density peakedness [ne(0)/〈ne〉] during the beam pulse. To date, the best fusion resu...

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, ...

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980`s The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America`s first operational D-T fusion reactor. This paper will discuss negative pressure employing `elephant trunks` in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and {Delta} pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs.

The Tokamak Fusion Test Reactor (TFTR) operated for 15 years, from 1982 to 1997. From November 1993 to April 1997 a mixture of deuterium-tritium (D-T) was used to fuel experiments, leaving the reactor system in an activated and contaminated state. In this condition, the removal of the TFTR and its subsystems is greatly complicated. In 1999, PPPL commenced the TFTR Decontamination and Decommissioning (D&D) Project. The objectives of the D&D projects are to completely remove the TFTR in a safe, efficient and cost effective manner. To date all diagnostics, structural components and ancillary systems have been removed, and the tokamak is in the process of being sectioned into smaller pieces. This paper will describe the criteria, the methods employed and the successes of the D&D Project.

The original design of the Tokamak Fusion Test Reactor (TFTR) and tritium handling are described. Modifications to these designs inspired by operational experience gained from running TFTR since 1982 with deuterium-deuterium (D-D)-fueled experiments are also presented. The presently installed gas handling systems work well for the D-D phase of operation, but they do not have the level of sophistication needed for tritium injection, tritium-contaminated-gas exhaustion, and gas detritiation for deuterium-tritium (D-T) operations. The pumping and gas-handling systems were designed prior to the operation of TFTR to handle the tritiated process streams anticipated during the DT phase. Since then, it has been found that certain changes and refinements to this design would make tritium handling simpler and safer. Most (90%) of the process gas (deuterium fuel) is pumped by the cryopanels in the neutral beams during operations. This nearly pure elemental hydrogenic gas is radiologically 25000 times more safe to handle than the corresponding amount of tritium oxide. It has been decided to transfer the hydrogenic species from the neutral beam cryopanels into special containers for off-site shipment as elemental tritium rather than as oxide. This reduces the exposure of oil-lubricated pumps from large contact with tritium-laden gas and reduces the risk of exposure to tritium oxide. Conditioning of the torus will be processed by impurity cryopumps, designed to pump only the nonhydrogenic species (impurities) and reduce the total amount of tritium used during conditioning. >

TFTR has been using tritium as a fuel gas as part of its Deuterium-Tritium (DT) experimental operations with over 600 kCi processed through the site and over 400 kCi injected. Careful inventory and accounting measures are required to ensure site and regulatory safety tritium limits are not exceeded. Tritium management is accomplished through a database system to coordinate experimental needs within these limits. A TFTR Nuclear Materials Custodian (NMC) oversees Tritium Transfer Operations (TTO) using accounts based on specific system physical locations including tritium product containers, uranium beds (hydration), Gas Holding Tanks (GHT), disposable Molecular Sieve Beds (DMSB), the Tritium Gas Injection System (TGIS), and the Tritium Purification System (TPS). Methods used to coordinate and track TTO's include: single source TTO serial numbers issued by the TFTR Shift Supervisor (TFTRSS) as TTO authorization, TTO verification receipts, and daily NMC database updates with reports issued to appropriate groups. Tokamak tritium operations are controlled and monitored through a computer system that automatically measures and generates tritium injection value reports, as well as providing interlocks between the TGIS and the personnel safety tritium monitors. This system also provides historical trends of tritium values in the entire TGIS, consisting of 14 injector volumes and an interconnecting manifold, to identify inter-system tritium leakage. NMC tritium accounting control involves aspects such as: tritium measurement data flow, verification, instrument calibration, measurement controls, TTO control limits, Material-in-Process (MIP) evaluation, database error detection and correction, and account material balances. Details of these areas are discussed with emphasis on the tight controls required for the very low tritium inventory level permitted for TFTR, and lessons learned from the past 23 months of tritium operations.

At the beginning of the 1990 Tokamak Fusion Test Reactor (TFTR) experimental run, after replacement of POCO‐AXF‐5Q graphite tiles on the midplane of the bumper limiter by carbon fiber composite (CFC) tiles and prior to any pulse discharge cleaning (PDC), boronization was performed. Boronization is the deposition of a layer of boron and carbon on the vacuum vessel inner surface by a glow discharge in a diborane, methane, and helium mixture. The amount of discharge cleaning required after boronization was substantially reduced compared to that which was needed after previous openings when boronization was not done. Previously, after a major shutdown, about 105 low current (∼20 kA) Taylor discharge cleaning (TDC) pulses were required before high current (∼400 kA) aggressive pulse discharge cleaning (PDC) pulses could be performed successfully. Aggressive PDC is used to heat the limiters from the vessel bakeout temperature of 150 to 250 °C for a period of several hours. Heating the limiters is important to in...

We routinely use quadrupole mass spectrometers (QMS) to monitor vacuum conditions, gas purity, and plasma–wall interactions in the Tokamak Fusion Test Reactor (TFTR) at Princeton. Two QMS systems have been operating on TFTR continuously for a 2 year period. Both QMS systems are absolutely calibrated at weekly intervals using a six‐part standard gas mixture. The calibration procedure is based on the use of transfer standards (ion gauge and capacitance manometer) that are calibrated against a primary standard (spinning rotor gauge) on an external vacuum system. We have identified variations in the efficiency of the QMS ionizer and drifts in the sensitivity of the electron multiplier ion detector to be the major reasons for the observed changes in overall QMS sensitivity. Weekly variations in sensitivity greater than 100% have been observed following system bakeout at 150 °C and with the use of rhenium filaments which were initially in the QMS ionizer. Operation of the QMS system with tungsten filaments and ...

A tritium compatible system has been designed to handle the exhaust gas produced by neutral beam operations in the Tokamak Fusion Test Reactor (TFTR). The system transfers gas highly concentrated in hydrogen isotopes to intermediate storage and final shipping containers using a combination of dry pumping and gettering. Details are given for functional and design requirements, component characteristics, and estimated system performance. Process flow and equipment arrangements are diagrammed, and various modes of operation are illustrated. The system is designed to provide safe handling of the tritiated exhaust and the rapid tritium turnaround necessary for operation with TFTR's small on-site inventory.

During the 1987 run, TFTR reached record values of QDD, neutron source strength, and Ti(0). Good confinement together with intense auxiliary heating has resulted in a plasma pressure greater than 3*105 Pascals on axis, which is at the ballooning stability boundary. At the same time improved diagnostics, especially ion temperature profile measurements, have led to increased understanding of tokamak confinement physics. Ion temperature profiles are much more peaked than previously thought, implying that ion thermal diffusivity, even in high ion temperature supershot plasmas, is greater than electron thermal diffusivity. Based on studies of the effect of beam orientation on plasma performance, one of the four neutral beamlines has been re-oriented from injecting co-parallel to counter parallel, which will increase the available balanced neutral injection power from 14 MW to 27 MW. With this increase in balanced beam power, and the addition of 7 MW of ICRF power it is planned to increase the present equivalant QDT of 0.25 to close to break-even conditions in the coming run.

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 Tokamak Fusion Test Reactor (TFTR) is scheduled to begin DT operation in 1990 with the on-site tritium inventory limited to 5 grams. The physics and chemistry of the in-vessel tritium inventory will impact safety concerns, and also the entire operating schedule of the tokamak. We have investigated plasma-material interaction processes that will affect this first tritium-fueled tokamak. Tritium inventory estimates for TFTR are derived from: (1) laboratory simulation, (2) in-situ plasma measurements, (3) post-run surface analysis, and (4) modeling. This paper presents the results of these investigations, the derivation of a tritium inventory estimate and its uncertainties, and a discussion of its impact. A particular discharge-by-discharge operating schedule has been developed and evaluated. The major source of in-vessel tritium inventory will be codeposition of tritium and eroded carbon onto surfaces. We find that the on-site limit may be approached unless specific inventory reduction techniques are invoked, e.g., discharge cleaning.

The principal heating system for the Tokamak Fusion Test Reactor (TFTR) consists of four beamlines to inject beams of neutral deuterium into the tokamak plasma. We have recently injected beams at total powers up to about 30 MW, and we have achieved ion temperatures in the plasma core of 30 keV and more.

A strong pumping effect was observed with plasma operation on the toroidal graphite bumper limiter in TFTR. The pumping effect was induced by conditioning the limiter with a short series (10-20) of low density deuterium or helium initiated discharges. The decay constant for gas fuelled Ohmic discharges was reduced from before conditioning to a minimum of after conditioning, which corresponds to a reduction in the global recycling coefficient from about 100% to less than 50%. Coincident with the low recycling conditions, low current neutral beam fuelled discharges showed global energy confinement times which were enhanced by a factor of two over the values obtained with an unconditioned limiter. Two models are proposed for the observed pumping effects: (1) a depletion model, based on pumping of hydrogenic species in the near-surface region of the limiter after depletion of the normally saturated surface layer by (carbon and helium) ion induced desorption; and (2) a co-deposition model, based on adsorption of hydrogenic species in carbon films formed by material sputtered from the limiter during conditioning.

We have studied the effects of the deposition of thin films (∼0.1 μm) of chromium onto ∼70% of the torus area of the Tokamak Fusion Test Reactor (TFTR). The purpose of these experiments was to test the difference between high surface coverage and high pumping speed gettering schemes with respect to minimizing oxygen impurity generation in high power tokamak discharges. The initial Cr deposition had significant effects on vessel outgassing and subsequent plasma performance: the outgassing of H2O, CO, and CO2 decreased by a factor of 10, oxygen impurity radiation decreased by a factor of 2, the plasma Zeff decreased from 1.3 to 1.1, and the plasma density limit increased by 20%. This improvement correlates with a significant reduction of the edge radiation as the density limit is approached. The effects of the initial and subsequent Cr depositions were relatively long lasting, exhibiting time constants of the order of weeks. We attribute the observed impurity reduction to a modification of the oxide surface...

The design and operation of instrumentation included in the gas injection system of the Tokamak Fusion Test Reactor (TFTR) for torus pressure and gas flow measurements are described. Magnetically shielded ion gauges, located on the torus boundary, are used for fast (1 kHz) torus pressure measurements over the range 10−7–10−3 Torr. Gas injection assemblies comprising an array of piezoelectric gas injection valves are situated at three toroidal locations. The gas injection valves are programmed for various discharge conditions including feedback control from the torus pressure gauges and plasma density measurement. Gas flow through the injection valves is monitored by strain‐gauge‐type pressure transducers on each injection valve. The resulting gas flow measurements are used for gas fueling, particle balance, and recycling studies.

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.