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It now appears feasible to deposit positrons (e+) in a tokamak plasma by injecting bursts of neutral positronium atoms (e+e−), which are then ionized by the plasma. The annihilation time of these positrons in the plasma is long compared with typical particle containment times. Thus the subsequent transport of the positrons can be studied by monitoring the time dependence of the annihilation, gamma radiation produced when the positrons strike a limiter. This paper discusses the design of such an experiment, the kinds of data which can be obtained, and the physics questions which this experiment might address. This diagnostic technique could also be useful in studying transport in other magnetic confinement devices such as reversed-field pinches and magnetic mirrors.

Large improvements in spheromak parameters and new understanding have been obtained from the CTX experiment at Los Alamos [Phys. Rev. Lett. 51, 39 (1983); 61, 2457 (1988)]. In one experiment the global energy confinement time has been increased an order of magnitude over previous experiments to 0.2 msec and the magnetic-energy decay time increased to 2 msec. These results were achieved in a decaying spheromak by reducing the helicity dissipation in the edge. In another smaller spheromak, record electron temperatures (∼400 eV) and record magnetic field strengths (∼30 kG) have been obtained.

An increase in the global energy confinement time (TE) was obtained in the CTX spheromak by replacing the high-field-error mesh-wall flux conserver with a low-field-error solid-wall flux conserver. The maximum TE is now 0.18 ms, an order of magnitude greater than previously reported values of ;S0.017 ms. Both TE and the magnetic energy decay time ixw) now increase with central electron temperature, which was not previously observed. These new results are consistent with a previously proposed energyloss mechanism associated with high edge helicity dissipation.PACS numbers: 52.55.Hc, 52.70.Kz A spheromak 1 is a toroidal magnetic configuration with large internal plasma currents and self-generated internal magnetic fields. This configuration has been studied for many years with the hope that it would make an attractive fusion reactor. However, an apparent condemning feature of the spheromak for reactor use was the short global energy confinement times (T^) previously reported, 2,3 in the 5-20-jis range. It has been proposed that the dominant energy loss has been a consequence of enhanced helicity dissipation in the edge region of the spheromak induced by magnetic-field errors. 2 " 4 (Helicity 5 K is the quantitative measure of the "knottedness" of magnetic-field lines, i.e., flux linkage.) In the case of CTX with a mesh-wall flux conserver, 2,6 the field errors were due to the bridges crossing the midplane gap, the coarseness of the mesh, and the nonzero resistivity of the copper rods. Edge field lines either contacted the rods themselves, or the vacuum tank surrounding the flux conserver. It was estimated that approximately 25% of the poloidal flux intersected metal, and it appears that the resistivity of the open field lines was dominated by electron-neutral collisions 2 (rj e . n is much higher than r/spitzer in this region). These field lines are influenced by the minimum-energy principle, 7 which states that ^=//oJ' B/| B | 2 (where J is the current density and B is the magnetic field) should be a spatial constant. Therefore, current is driven on the highresistance open field lines, primarily by instabilities in the bulk plasma induced by Vk. The exact nature of these instabilities is not yet understood, but it is generally accepted that they cause direct ion heating at the expense of magnetic energy (W), giving 2,3,6 ion temperatures (Tj) higher than the electron temperature (T e ). The enhanced decay rate of W (due to instabilities and direct ion heating) must be accompanied by an approximately equal helicity decay rate (T/T'= -K/K), because W/Koc{X) (volume-averaged A.), which changes only slightly with VA. Enhanced AT is a direct result of large edge 77J, since 8 Koc J V0 \T]J' Bd 3 x. (One can think of this as enhanced "untying" of the "magnetic knot" in the resistive edge.) The severe impact on x E came from high charge-exchange rates involving the hot, directly

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Recent improvements in the operation of the CTX spheromak device have produced discharges containing evidence for a pressure-driven instability. The instability leads to a distinct event in the discharge, which can be studied in detail. Data are presented which reasonably discount Taylor relaxation of the current profile as the cause of the event. The critical value of the normalized pressure gradient has been measured and is compared with the Mercier limit.PACS numbers: 52.55.Hc, 52.35.Py A spheromak is a toroidal magnetic configuration with large toroidal and poloidal plasma currents which generate most of the internal magnetic field. This configuration is attractive for a fusion reactor because it is compact, has a high level of Ohmic heating power, and thus, may not require auxiliary heating (such as neutral beam or rf injection), and has a high engineering p (pcng&P/Blaw). However, the theoretically predicted internal p (p yo \ocP/(B 2 \ 0 \) calculated with the Mercier criterion is small, ranging from « 0.2% for the "classical" spheromak 1 to (1-2)% for nonspherical cross sections, 1 " 3 to ?£7% for the spheromaks with current holes. 2 " 4 In fact, many other toroidal magnetic configurations have low predicted p limits. Tokamaks are theoretically limited by the ballooning criterion 5 to ;S(5-10)%, and reversed field pinch equilibria stable to current-driven resistive tearing modes have a predicted p limit 6 of « 20%. The experimental significance of predicted p limits is often unclear, since other effects might limit plasma performance. If the p limits could be reached, the expected plasma behavior is often unknown. Tokamaks historically have had difficulty reaching the predicted p limit, 5 but recently large tokamaks have been used to explore plasma conditions at reactor relevant p values. The highest values obtained are empirically found to follow a scaling law of the form Pmax^3I/aB On %, MA, m, and T) 7,8 ; however, this can be consistent with either ballooning mode or kink mode limits. 7 In addition, operation at this p limit is found to be disruptive 7 in some cases and in other cases a degradation of energy confinement without disruption is observed. 8 Reversed field pinches normally operate 6 with roughly constant p («10%), indicating an empirical limit, but the value is less than the predicted value, which does not include the localized resistive interchange mode (sometimes called the "g mode"). In contrast, spheromaks often have p values above those predicted. 9 ' 10 In this Letter, strong evidence for a pressure-driven instability in CTX 11 spheromaks is presented. The instability leads to a distinct event in the discharge which can be analyzed in detail. It is found that when a particular threshold value in the pressure gradient is exceeded, internal plasma is expelled toward the wall in 10-20 jus. The resulting temperature and density profiles are both hollow (about the magnetic axis), strongly indicating a magnetic flux interchange. These hollow profiles are short lived, and evolve back to mod...

Alcator C operations commenced with discharge cleaning and tokamak operation using hydrogen filling gas.Prior to and during these experiments no deuterium gas was allowed into the device. The earliest operation resulted in dosimeter readings of a few Roentgen per shot in the vicinity of the limiter and a localized source of neutron emission of up to 109 neutrons per shot which were subsequently identified as having photonuclear origin. After seven months of operation, conditions were achieved that resulted in substantially less photonuclear activity. Subsequently, deuterium fill gas was allowed into the device and measurement of neutron flux and energy spectra indicated that the majority of neutron emissions in Alcator C high density deuterium discharges were consistent with having thermonuclear origins.

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