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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

Boronization of NSTX has allowed access to higher density higher confinement plasmas. A glow discharge with 4 mtorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B + C areal density of 1018 cm-2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by a factor of 15, carbon emission reduced by a factor of two and copper reduced to undetectable levels. After boronization, the plasma current flat-top time increased by 70% enabling access to higher density higher confinement plasmas.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

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.

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

Recent experiments on the Tokamak Fusion Test Reactor have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition, and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) transiently increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean‐up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing ≊8000 Ci and restoring the tritium inventory to a level well below the administrative limit.

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.

There have been many significant changes in the status of tritium activities in the US since the 4th Tritium Conference in October, 1991. The replacement Tritium Facility (RTF) at Savannah River Site and the Weapons Engineering Tritium Facility (WETF) at the Los Alamos National Laboratory are now operational with tritium. The Tokamak Fusion Test Reactor (TFTR) has initiated a highly successful experimental campaign studying DT plasmas, and has produced more than 10 Megawatts (MW) of fusion power in a D-T plasma. Sandia National Laboratory has ceased tritium operations at the Tritium Research Laboratory (TRL) and many of the activities previously performed there have been transferred to Los Alamos and Savannah River. The tritium laboratory at Lawrence Livermore National Laboratory has reduced the tritium inventory to

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.

Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 1017 m ₋3 without the appearance of either disruptive magnetohydrodynamics events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits.

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

A number of important applications of ultrahigh vacuum technology involve the use of vacuum systems which cannot be baked in situ, e.g., materials processing equipment and various designs of accelerators and fusion devices. A choice of vessel material and surface treatment which minimizes outgassing in unbaked systems has obvious advantages. We report on a systematic comparison of outgassing measurements from type 304 stainless steels and type 6061/63 aluminum alloys after various standard and advanced surface treatments. Correlations were made with high‐resolution secondary ion mass spectroscopy and x‐ray photoelectron spectroscopy profiles of the passivation oxide layer and profilometry measurements of the surface roughness. The observed outgassing rates Q obeyed a power law Q=Q 0 t −n where n varied from 0.94 to 1.27. Over the standard observation period (50 h), the outgassing was dominated (≳85%) by H2O. The absolute outgassing rates varied by only a factor of 2.8 for the stainless steels and by a factor of 1.4 for the aluminum alloys, even though the surface roughness differed by more than two orders of magnitude. The high‐value‐added surface treatments (i.e., electropolishing) did not have significantly lower outgassing in comparison to simple detergent washes.

The vacuum pumping requirements for a fusion device are dictated by a number of factors, including the materials used in construction of the device, the cleaning and conditioning techniques implemented, the operating conditions and device temperature, the plasma characteristics, and the fuel gases and impurities retained and released by the plasma-facing components (PFCs). In an attempt to derive guidelines for determining the vacuum pumping requirements of a generic fusion device, a study was undertaken to examine the vacuum pumping capabilities of existing large fusion devices, to review the cleaning and conditioning techniques now in use, and to catalog pertinent vacuum equipment that is now available or anticipated soon. In a survey of six large fusion devices [ASDEX, DIII-D, the Joint European Torus (JET), JT-60, the Tokamak Fusion Test Reactor (TFTR), Tore Supra], information was collected on PFC materials, cleaning and conditioning techniques, device operating temperatures, and vacuum pumping system characteristics and on the surface conditions and partial pressures necessary to achieve high-purity plasma discharges in each device. The results of the survey are reviewed to determine how the various factors affect vacuum pumping requirements. The use of tritium in future fusion devices is expected to create a new set of problems for the vacuum pumping systems of these devices. These problems are addressed, and some possible solutions for the pumping and gas handling systems are identified. New vacuum equipment is under development, and the fusion community is in a position to influence its direction and emphasis. Information gained in studies of this type should be brought to bear on the development process to ensure that vacuum equipment will be available to meet the needs of future fusion devices.

Recent research on TFTR has emphasized optimization of performance in deuterium plasmas, transport studies and studies of energetic ion and fusion product physics in preparation for the D-T experiments that will commence in July of 1993. TFTR has achieved full hardware design parameters, and the best TFTR discharges in deuterium are projected to QDT of 0.3 to 0.5.The physics phenomena that will be studied during the D-T phase will include: tritium particle confinement and fueling, ICRF heating with tritium, species scaling with tritium, collective alpha-particle instabilities, alpha heating of the plasma and helium ash buildup. It is important for the fusion program that these physics issues be addressed to identify regimes of benign alpha behavior, and to develop techniques to actively stabilize or control instabilities driver by collective alpha effects.

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...

A plasma-enhanced CVD process was used to deposit thin (40–100 nm), amorphous carbon/boron films onto the first wall of TFTR. A series of plasma experiments was performed to test the effect of this first-wall modification on low-Z impurity (carbon, oxygen) behavior, suppression of metallic impurities, post-disruption recovery, hydrogenic recycling, and plasma density limits. The analysis of the plasma performance following the first few boronization attempts on TFTR shows: (1) suppression by a factor of 3-1.5 of carbon impurities immediately after boronization, although this suppression is short-lived (~10 shots); (2) long term suppression of oxygen and metallic impurities by a factor of two; (3) significant suppression of oxygen impurity emission following disruptions and during pulse discharge cleaning; and (4) no significant change in hydrogenic recycling properties in comparison to the pre-existing carbon first wall.

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 TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.

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.

Boronization on the National Spherical Torus Experiment (NSTX) has proved to be quite beneficial with increases in confinement and density, and decreases in impurities observed in the plasma. The boron has been applied to the interior surfaces of NSTX, about every 2 to 3 weeks of plasma operation, by producing a glow discharge in the vacuum vessel using deuterated trimethylboron (TMB) in a 10% mixture with helium. Special NSTX requirements restricted the selection of the candidate boronization method to the use of deuterated boron compounds. Deuterated TMB met these requirements, but is a hazardous gas and special care in the execution of the boronization process is required. This paper describes the existing GDC, Gas Injection, and Torus Vacuum Pumping System hardware used for this process, the glow discharge process, and the automated control system that allows for remote operation to maximize both the safety and efficacy of applying the boron coating. The administrative requirements and the detailed procedure for the setup, operation and shutdown of the process are also described.

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.

Maintenance and associated line breaks are a requirement for any operating vacuum systems. When the vacuum systems are potentially contaminated with tritium, the maintenance crews are faced with opening a line that may contain significant quantities of airborne tritium. Every isolatable section of complex vacuum systems cannot contain an active tritium monitor, therefore, an alternate method of tritium measurement must be used in preparation for a line break. At TFTR, the method used to determine the quantity of tritium at risk prior to a line break is accomplished by Operations Procedure OP-VAC-656 (Approved Method for Venting and Pumping a Vacuum Space That Has Possible Tritium Contamination). This procedure describes the method to safely vent a vacuum space that has been exposed to possible contamination with tritium, to measure the concentration of airborne tritium and to provide a set of options for pumping and purging based on the tritium concentration. The procedure describes the minimum compliment of components needed to safely perform the process. At the present time at TFTR there are at least four different sets of components that meet the requirements of OP-VAC-656. Each of these special setups are used because of the unique system requirements in the different areas of TFTR.

Boronization of the National Spherical Torus Experiment (NSTX) has enabled access to higher density, higher confinement plasmas. A glow discharge with 4 mTorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B+C areal density of 10 to the 18 (B+C) cm to the -2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by x15, carbon emission reduced by two and copper reduced to undetectable levels. After boronization, the plasma current flattop time increased by 70% enabling access to higher density, higher confinement plasmas.

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...

The present study applies the approach described in Mark et al. for the testing of toxic chemicals produced during the drying of Douglas-fir. The genotoxic potential of Douglas-fir condensate has been previously unexplored and is thus an area of appropriate concern to the forest products industry, regulatory, agencies, and the general public. Previous research conducted in this laboratory has identified two wood-drying condensates that yield positive cytotoxic and genotoxic effects. The results of testing Southern yellow pine and Eastern white pine condensates have been reported elsewhere. Douglas-fir condensate, a third wood-drying condensate, was added in vitro in concentrations ranging from 0.1 to 100 microliters/ml to cultures of Chinese Hamster Ovary (CHO-WBL) cells. A dose response curve was observed with this condensate for both cytotoxicity and genotoxicity. The number of viable cells as well as the mitotic index (MI) and proliferative rate index (PRI) varied inversely with dosage. The result of chromosome aberration (Abs) analysis and sister chromatid exchange (SCE) analysis, both cytogenetic measures of genotoxicity, also gave statistically significant results.

Eastern white pine is one of the most important commercial species of wood in the Northeast. Condensates extracted from this wood were tested to detect potential cytotoxicity and genotoxicity in Chinese Hamster Ovary (CHO) cells in the absence of S-9 activation. Cytotoxicity was measured by the Trypan blue exclusion assay, mitotic index (MI) and proliferative rate index (PRI). Genotoxicity was measured by the chromosome aberration (CA) assay and sister chromatid exchange (SCE) analysis. Both cytotoxic and genotoxic effects were observed. Laboratory-generated Eastern white pine condensate reduced the viability of CHO cells. The number of viable cells was roughly inversely proportional to dosage over a range of 91 percent to 58 percent survival in treated groups as compared to 2.4 x 10(5) viable cells (100 percent) in the control. The mitotic index (MI) data also showed an inverse correlation with dosage. The highest scorable dose limited by toxicity was determined to be 1 ml of Eastern white pine condensate in a total of 10 ml of medium. Lastly, a dose response curve was observed using the CA assay and also with the SCE analysis. The present findings support results obtained from Ames testing of Eastern white pine condensate and also corroborate results derived from human peripheral-blood lymphocytes.

A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor (TFTR) in deuterium plasmas heated by co and counter injected deuterium and tritium neutral beams with a total power of 33.7 MW. The ratio of fusion power output to heating power input is 0.27. At the time of the highest neutron flux the plasma conditions are: Te(0)=11.5 keV, Ti(0)=44 keV, ne(0)=8.5×1019 m−3, and 〈Zeff〉=2.2 giving τE=0.24 s. These conditions are similar to those found in the highest confinement deuterium plasmas. The measured D‐T neutron yield is within 7% of computer code estimates based on profile measurements and within experimental uncertainties. These plasmas have an inferred central fusion alpha fraction of 0.2% and central fusion power density of 2 MW/m3 similar to that expected in a fusion reactor. Even though the alpha velocity exceeds the Alfven velocity throughout the time of high neutron output in most high power plasmas, MHD activity is not substantially different from that in co...

Condensate from eastern white pine, one of the commercially most important species of tree in the northeastern United States, was treated for potential cytotoxicity and genotoxicity in human peripheral blood lymphocytes in the absence of S-9 activation. Cytotoxicity was evaluated by mitotic index (MI) determination and proliferative rate index. Genotoxicity was measured by the chromosome aberration (CA) assay and sister chromatid exchange (SCE) analysis. Both cytotoxic and genotoxic effects were observed with laboratory-generated eastern white pine condensate. The MI data showed an inverse correlation between the MI and treatment dosage. A dose response curve was observed using the CA assay and also with the SCE analysis. The present findings thus corroborate the results from Ames testing and represent the only information currently available on the cytotoxic and genotoxic potential of these chemicals.

A major activity of the lumber industry is the kiln-drying of wood. In order to ascertain whether wood-drying condensates pose a possible environmental hazard, the cytotoxicity and genotoxicity of these condensates in vitro, were tested using an assay validated using Chinese hamster ovary (CHO) cells and a known genotoxicant, mitomycin C. Subsequently, the assay was developed for the human peripheral blood lymphocyte (HPBL) system, as it was felt that results derived from human cells would reflect the situation more closely in vivo. Condensates from Southern yellow pine, Eastern white pine and Douglas fir trees were tested in CHO and HPBL systems and have demonstrated cytotoxic and genotoxic effects in vitro, as reported elsewhere. Red oak condensate has also been tested using the HPBL system. Thus far, results are consistent with the hypothesis that there is no difference between the cytotoxic and genotoxic effects of treated cells versus controls. This finding indicates either that the condensate of red oak poses no appreciable genetic hazard as measured by cytotoxicity and genotoxicity assays, or that the condensate has lost its potency with time and storage; both of these possibilities have important environmental implications.

The Tokamak Fusion Test Reactor (TFTR) has performed initial high-power experiments with the plasma fueled by deuterium and tritium to nominally equal densities. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by ~20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by α-particles.

We describe the design for the vacuum vessel and vacuum pumping system for a 1.2 GeV electron storage ring for light source applications. The hybrid vacuum vessel design consists of eight extruded aluminum (6063‐T4) dipole chambers curved to the ring radius of 2.93 m, with interconnecting stainless (304 L) steel straight vessel segments. The dipole chambers contain two 60 milliradian photon exit ports on the outer midplane, and an integral pumping chamber on the inner midplane which provides 500 1/s ⋅ m of active gas pumping with a circumferential array of a non‐evaporable getter (NEG) material. Ion pumping stations are distributed along the straight vacuum vessel segments to provide for pumping of CH4 and noble gases. The distributed pumping provided by the NEG offers a factor of five to ten times larger pumping speed in comparison to conventional designs based solely on the use of discrete or distributed ion pumps. Externally bonded cooling tubes on the dipole chambers offer a cost effective and fault tolerant means of removing the x‐ray heat load. Estimates of the conditioning time for the vacuum vessel are presented based on the recent photodesorption data from aluminum surfaces.

It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: (1) Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure, and (2) aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a){approx}3 at {approx}250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H{sub 2}O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of X-rays from runaway electrons, is an indication that TDC is no longer necessary after {approx}10{sup 5} pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 {degree}Cmore » to about 250 {degree}C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H{sub 2}O is removed from the graphite limiter.« less

The plasma contacting surfaces of the TFTR movable limiter consist of an array of titanium carbide coated graphite tiles. The surface characteristics of the graphite tiles were analyzed prior to installation to establish the initial limiter conditions, and after exposure to 800 Ohmically heated discharges in TFTR to provide data on the limiter conditioning process. Some tile surfaces showed nearly complete removal of the TiC surface as a result of plasma exposure. 3D and 2D elemental mapping of the tile near surface regions show deposition of vacuum vessel metals (primarily Ti and Ni) in the form of oxides.

The duration of meiosis and spermiogenesis was studied in the dog by following the kinetics of [3H]thymidine-labeled canine spermatocytes autoradiographically. Leptotene appears to last for at least 4.20 days, zygotene for less than a day, and pachytene for about 15.50 days. The transition from diplotene to metaphase II is extremely rapid and requires approximately a day. The duration of spermiogenesis is at least 19.60 days. Both meiosis and spermiogenesis are completed by 42.15 days.

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 ...

The conditioning techniques applied to the TFTR first-wall configuration that will be in place for the DT experiments in 1990–1991 are reviewed. Of primary interest is the helium conditioning procedure that was developed to control hydrogenic recycling from the graphite, inner-wall bumper limiter. Operation of TFTR over the plasma density range for gas-fueled ohmic plasmas, ne = (2–5) × 1019m−3, typically results in hydrogenic recycling coefficients near unity. The use of the helium conditioning procedure produced recycling coefficients as low as 0.5, and decreased the minimum ohmic plasma density to ne = 0.5 × 1019m−3 at IP = 0.8 MA. Low density ohmic target plasmas with low recycling conditions are prerequisite conditions for the enhanced confinement (e.g., “supershot”), neutral-beam-heated discharges observed in TFTR during 1986–1987, which is the primary mode being considered for study in the DT experiments. The recycling changes induced by the helium conditioning procedure are believed to be the result of a plasma pumping effect in the graphite induced by He and C ion desorption of hydrogenic species from the near-surface (< 20 nm) layer of the limiter. The capacity of the conditioned limiter to pump gas-fueled, pellet-fueled, and neutral-beam-fueled discharges is compared. The helium conditioning technique is also beneficial for isotopic exchange and for minimizing the in-vessel tritium inventory.

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.

We have developed a method for passivating previously activated ZrAl alloy bulk getters. Such passivation is essential for avoiding embrittlement of the getters during discharge cleaning operations in tokamak fusion devices. The method consists of exposing the getters (at room temperature) to a controlled amount of oxygen, followed by pulse discharge cleaning to restore the vacuum vessel walls to clean conditions. We have tested this method on the Tokamak Fusion Test Reactor (TFTR) and present measurements of the decrease in the hydrogen and oxygen pumping speeds as functions of the quantity of oxygen adsorbed. These measurements are in good agreement with previous laboratory results. Using this method followed by 20 h of pulse discharge cleaning, we have reduced the getter pumping speed by more than a factor of 2×103 and have recovered the same clean vacuum conditions as existed prior to the oxygen exposure.

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.

Magnetic fusion devices require fast‐responding, versatile, and accurate gas injection systems in order to initiate and modulate the plasma density. We describe the gas injection control system (GICS), which regulates the gas injected into the torus of Princeton’s Tokamak Fusion Test Reactor (TFTR). Synchronized with each TFTR discharge, a PDP 11/23 microcomputer provides real‐time control for one to six fast injection valves, each allocated to feedback or open loop operation. Up to four feedback sources, including torus pressure gauges and plasma density measurements, are checked once each cycle for 1000 cycles; the cycle duration can be set to any multiple of 2 ms up to 16 ms. Detection of various fault conditions can cause automatic shutdown or modification of GICS operation. Data is archived after each TFTR discharge, and subsequently used for particle fueling studies and analysis of system drifts. A steady state mode of operation is used for vacuum vessel discharge cleaning. The system also provides ...

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.

Isotopic exchange experiments performed in the TokamakFusion Test Reactor (TFTR) are used to examine the outgassing and diffusive properties of graphite used as the plasmalimiter. Changeover from hydrogen to deuterium for different periods ranges from ∼600 to 60 plasma discharges, which appears to be correlated in the limitertemperature. We present a simple analytical model that predicts a fast transient (∼10 plasma discharges) changeover, where the deuterium fueling dilutes the adsorbed and near surface hydrogen, and a slowly changing term where bulk hydrogen diffuses to the surface. Using this model we can extract an activation energy for diffusion of 0.15±0.02 eV. We hypothesize that interpore diffusion for this porous (∼15%) material is consistent with our observations.

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.