A transport code (TRANSP) is used to simulate future deuterium-tritium (DT) experiments in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modelling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modelling. Fusion energy yields and 01 particle parameters are calculated, including profiles of the 01 slowing down time, the 01 average energy, and the AlfvBn speed and frequency. Two types of simulation are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the Do in the neutral beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that 01 heating will enhance the plasma parameters or that confinement will increase with the addition of tritium. The maximum relative fusion yield calculated for these simulations is QDT-0.3, and the maximum a contribution to the central toroidal 0 is PJO)-0.5%. The stability of toroidicity induced Alfvkn eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the simulations, particularly after the termination of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed.
Recent calculations have shown that when external momentum sources and plasma rotation are included in the neoclassical theory, the standard results for impurity transport can be strongly altered. Under appropriate conditions, inward convection is reduced by co-injection and enhanced by counter-injection. In order to examine the theoretical predictions, several observations of impurity transport have been made in the ISX-B tokamak during neutral-beam injection for comparison with the transport seen with Ohmic heating alone. Both intrinsic contaminants and deliberately introduced test impurities display a behaviour that is in qualitative agreement with the predicted beam-driven effects. These correlations are particularly noticeable when the comparisons are made for deuterium where the impurity transport in the Ohmically heated discharges exhibits neoclassical-like characteristics, i.e. accumulation and long confinement times. Similar but smaller effects are observed in beam-heated hydrogen discharges; neoclassical-like behaviour is not seen in Ohmically heated hydrogen sequences. Emphasis has been placed on measuring toroidal plasma rotation, and semiquantitative comparisons with the theories of beam-induced impurity transport have been made. It is possible that radial electric fields other than those associated with momentum transfer and increased anomalous processes during injection could also play a role.
Neutral-beam injection of up to 2.5 MW into plasmas in the ISX-B tokamak (R0 = 0.93 m, a = 0.27 m, BT = 0.9–1.5 T, Ip = 70–210 kA, n̄e = 2.5–10×1013 cm−3) has created plasmas with volume-averaged beta of up to ∼ 2.5%, peak beta values of up to ∼ 9%, and root-mean-square beta values of up to ∼ 3.5%. Energy confinement time is observed to decrease by about a factor of two as beam power goes from 0 to 2.5 MW; the decrease is caused predominantly by the electron confinement time falling below the predictions of ‘Alcator scaling’ by a factor of 3–4 at high beam power. An empirical relationship of the form fits our measurements over a wide range of plasma parameters. The function f(Pb), where Pb is the beam power, is linear for Pb ≤ 1.2 MW but tends to saturate for 1.2 MW ≤ Pb ≤ 2.5 MW. Although the equilibria attained in ISX-B are predicted to be above the threshold for the ideal magnetohydrodynamic (MHD) ballooning instability, no evidence of these modes is observed.
The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in the Tokamak Fusion Test Reactor [Plasma Phys. Controlled Fusion 26, 11 (1984)]. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. Radial profiles of the deuterium and tritium densities are determined from the D-T fusion neutron emission profile. The inferred scalings with average isotopic mass are quite
The measured D-D neutron rate of neutral beam heated JET baseline and hybrid H-modes in deuterium is found to be between approximately 50% and 100% of the neutron rate expected from the TRANSP code, depending on the plasma parameters. A number of candidate explanations for the shortfall, such as fuel dilution, errors in beam penetration and effectively available beam power have been excluded. As the neutron rate in JET is dominated by beamplasma interactions, the 'neutron deficit' may be caused by a yet unidentified form of fast particle redistribution. Modelling, which assumes fast particle transport to be responsible for the deficit, indicates that such redistribution would have to happen at time scales faster than both the slowing down time and the energy confinement time. Sawteeth and edge localised modes are found to make no significant contribution to the deficit. There is also no obvious correlation with magnetohydrodynamic activity measured using magnetic probes at the tokamak vessel walls. Modelling of fast particle orbits in the 3D fields of neoclassical tearing modes shows that realistically sized islands can only contribute a few percent to the deficit. In view of these results it appears unlikely that the neutron deficit results from a single physical process in the plasma.
Most auxiliary heating methods provide heating to more than one particle species (electrons, ions, impurities) in a fusion plasma. This can lead to substantial temperature differences between species, depending on conditions such as heating power to the different species and collisionality, with temperature differences between species limited by inter-species thermal equipartition and transport. The analysis of the steady-state electron-ion and impurity-ion power balances presented in this paper are used for consistency-checking experimental ion and electron temperature measurements and for inferring the main ion temperature from measured impurity temperatures. As ion temperature measurements by charge exchange spectroscopy (CXS) based on impurity ions have become more difficult and time-consuming since the installation of the ITER-like wall (ILW) with Be and W PFC’s, knowing the maximum sustainable temperature difference between ions and electrons, |Ti − Te| allows rejecting erroneous measurements. It also obviates the need for an ion temperature measurement, if an electron temperature measurement is available and |Ti − Te| cannot be larger than the combined errors of the underlying measurements. A power balance analysis is also required for estimating the errors of the ion and electron heat fluxes prior to any species-resolved transport analysis. The ion-impurity temperature differences are usually found to be small due to strong thermal equipartition between ion species. However, they can approach 10% in JET-ILW low density, high power discharges, such the ones under development for a future JET deuterium–tritium campaign (Joffrin et al 2019 Nucl. Fusion). This has a generally small, but not always negligible effect on the calculation of fusion reaction rates, which depend on main ion temperatures. An important outcome of this analysis is that temperature differences between impurity species are always much smaller than between the impurities and hydrogenic species and can usually be neglected. The paper presents two methods for calculating the impurity-to-main ion temperature ratio. Finally, this analysis leads to a method for the reconstruction ion temperature profiles from ion temperature data available at only one or a small number of spatial locations.
Short duration (20 msec) neutral deuterium beams are injected into the TFTR tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. I, p. 5 11. The subsequent confinement, thermalization, and diffusion of the beam ions are studied with multichannel neutron and charge exchange diagnostics. The central fast-ion diffusion ( co.05 m2/sec) is an order of magnitude smaller than typical thermal transport coefficients.
Scalings of the central rotation in non-gettered, co-injected ISX-B discharges have been measured as a function of beam power, electron density and plasma current. Extensive studies are made possible by exploiting charge-exchange excitation (CXE) of 0 VIII lines to measure Doppler shifts. The rotation velocity, vϕ(0), tends to saturate at (1.0 − 1.2) × l07 cm·s−1 when Pb≅0.5 MW, showing little further increase up to the maximum input of 2 MW; vϕ(0) is independent of ne and Ip. Momentum confinement times in quasi-steady plasmas are 10–16 ms for n̄e = 4.5 × 1013 cm−3. Counter-injection discharges always disrupt, but before this event vϕ(0) is the same as for co-injection plasmas. The addition of a third beam line, permitting injection of up to 2 MW of balanced neutral-beam power, has allowed comparisons of the energy and particle confinement in rotating and non-rotating plasmas with the same total neutral-beam input. In those cases where impurity buildup can be avoided, it is found that the ISX-B empirical scaling of energy confinement time is reproduced with balanced injection. Thus, the unfavourable dependence of is not the result of rotation. Studies of impurity behaviour under differing injection conditions have been extended to include fully stripped low-Z ions. The results are consistent with previous investigations of metallic elements which revealed strong dependences on the sense (co versus counter) of injection. The potentials calculated from momentum balance, using measured rotation profiles and typical plasma density and temperature profiles, are in qualitative agreement with the potentials measured directly for various combinations of co- and counter-injection.
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