Current Ramp-down Research Articles

The scaling of fuel recovered following un-mitigated disruptions in Alcator C-Mod with high-Z PFCs

The retention of fuel in plasma facing components (PFCs) is important for the viability of fusion – both in terms of economics and safety. This study shows that a single, un-mitigated, disruption can remove >30× more fuel than that retained in a single, 1s, C-Mod discharge with molybdenum and tungsten PFCs. The fuel is recovered due to heating of the near-surface (∼100μm) during the thermal and current quench periods of the disruption. A regression analysis of full-current disruptions in a dataset of 3200 discharges leads to a scaling of fuel recovered approximately proportional to WTH1×WMAG2 where WTH and WMAG are the thermal and poloidal magnetic energy inside the vessel respectively. Scaling by surface area and disruption time scales to ITER indicate 5–10MA plasmas with low thermal energy (e.g. during current rampdown) may be ideal for removing fuel from plasma-wetted surfaces.

Identification of anomalous Doppler resonance effect during current ramp down in HT-7 tokamak

The abrupt steep jump of electron cyclotron emission (ECE) signals during current ramp-down has been observed and explained by an anomalous Doppler resonance effect (ADR). The identifying process of ADR was presented based on the fast Fourier transform (FFT) technique. The threshold value for triggering a steep jump on ECE signals has been identified under different discharge conditions.

Experimental simulation of ITER rampdown in DIII-D

Because of the high value of the stored energy in planned ITER plasmas, the safe and controlled termination of ITER discharges is an important aspect of ITER operation. In a series of experiments on the DIII-D tokamak, we have simulated the proposed ITER rampdown for the 15 MA baseline operating scenario. These scaled discharges match the reference scenario (including scaled time dependence) with regard to key parameters such as scaled current, poloidal beta, elongation and internal inductance. The scaled plasma current is reduced to the equivalent of less than 1 MA, well below the 1.4 MA specified for ITER as the maximum allowable for disruptive termination. The plasma shape and position are controlled during rampdown so that the high heat flux zones near the strike points of the separatrix are held within the equivalent of the armoured zones of the ITER divertor; the regulation of the strike-point location is an order of magnitude better than required. Scans of the current rampdown rates indicate that a more rapid rampdown than the ITER reference case may be needed to avoid excessive current in the ITER central solenoid. Rampdown with a full-size plasma was studied, but was found to be unsuitable for ITER because of transitions to ELM-free H-mode with a consequent lack of density control, as well as large excursions in poloidal beta and internal inductance. We find that ELMs play an important role during the H-mode phase of the rampdown, helping to reduce the density as the current is reduced.

Tritium recovery in ITER by radiative plasma terminations

Planned radiative plasma terminations are examined as a method to recover tritium from plasma-deposited layers in the ITER tokamak. The technique exploits the high energy density of the ITER plasma, which is converted into a quasi-uniform radiation pulse by massive impurity injection that benignly terminates the plasma discharge. The radiation pulse transiently heats all plasma-viewing surfaces in order to desorb the tritium, which is released into the vessel and recovered by pumping. Calculations indicate significant tritium removal at reduced plasma current, ∼6–10MA, indicating the possibility of routine T recovery during the current rampdown phase of each discharge or during low current tritium recovery discharges.

Predictive transport modelling of type I ELMy H-mode dynamics using a theory-motivated combined ballooning–peeling model

This paper discusses predictive transport simulations of the type I ELMy high confinement mode (H-mode) with a theory-motivated edge localized mode (ELM) model based on linear ballooning and peeling mode stability theory. In the model, a total mode amplitude is calculated as a sum of the individual mode amplitudes given by two separate linear differential equations for the ballooning and peeling mode amplitudes. The ballooning and peeling mode growth rates are represented by mutually analogous terms, which differ from zero upon the violation of a critical pressure gradient and an analytical peeling mode stability criterion, respectively. The damping of the modes due to non-ideal magnetohydrodynamic effects is controlled by a term driving the mode amplitude towards the level of background fluctuations. Coupled to simulations with the JETTO transport code, the model qualitatively reproduces the experimental dynamics of type I ELMy H-mode, including an ELM frequency that increases with the external heating power. The dynamics of individual ELM cycles is studied. Each ELM is usually triggered by a ballooning mode instability. The ballooning phase of the ELM reduces the pressure gradient enough to make the plasma peeling unstable, whereby the ELM continues driven by the peeling mode instability, until the edge current density has been depleted to a stable level. Simulations with current ramp-up and ramp-down are studied as examples of situations in which pure peeling and pure ballooning mode ELMs, respectively, can be obtained. The sensitivity with respect to the ballooning and peeling mode growth rates is investigated. Some consideration is also given to an alternative formulation of the model as well as to a pure peeling model.

Analysis of an in-service rupture of the inner first wall of tore supra

The last experimental campaign of Tore Supra was also the last one with a set of plasma facing components mainly dating from the assembly of the machine in 1988. This campaign was interrupted a few days ahead of schedule because of the rupture of one coolant tube of the Inner First Wall (IFW). The rupture was observed by an abrupt increase of the pressure in the vessel. It was caused by a water leak which occurred 2 s after the end of the lower hybrid heating pulse during the current ramp-down. Investigations show that the accident is not related to the operation of the machine or a weaker element but rather to a localised heat flux exceeding the heat removal capability of the IFW.

Discharge optimization and the control of edge stability

Discharge optimization for improving MHD stability of both core and edge was essential for the achievement of record fusion power discharges, in the ELM-free hot ion H mode regime, in the recent JET DT operation. The techniques used to increase edge stability are described. In particular the successful technique of current rampdown used to suppress the outer mode is reported. The increased stability of the outer mode by decreasing the edge current density confirms its identification as an n = 1 external kink. Decreasing the plasma current, however, decreases the ELM-free period, which is consistent with stability calculations that show an earlier onset of the ballooning limit. In order to increase external kink stability without causing a deterioration in the ELM-free period, a compromise was achieved by using plasma current rampdown, while working at the highest plasma current values possible. Results from a plasma current scan show that at the time of occurrence of the first giant ELM, the plasma stored energy, as well as the pressure measured at the top of the edge pedestal increase linearly with plasma current, for a given plasma configuration and power. This is consistent with models of the edge transport barrier, where the transport barrier width is proportional to the ion (or fast ion) poloidal Larmor radius. The MHD observations in DT and deuterium only discharges were found to be similar. Thus the experience gained on the control of MHD modes in deuterium plasmas could be fully exploited in the DT campaign.

A study of maximum β in experiments with a current ramp-down in the TUMAN-3M tokamak

A study of maximum β in experiments with a current ramp-down in the TUMAN-3M tokamak

Experimental study of the -limit in ohmic H-mode in the TUMAN-3M tokamak

Because of its high confinement properties, the H-mode provides good opportunities to achieve high beta values in a tokamak. In this paper the results of an experimental study of and limits in the H-mode, obtained in a circular cross section tokamak without auxiliary heating are presented. The experiments were performed in the TUMAN-3M tokamak. The device has the following parameters: (limiter configuration), . The stored energy was measured using diamagnetic loops and compared with W calculated from kinetic data obtained by Thomson scattering and microwave interferometry. Measurements of the stored energy and of the were performed in the ohmic H-mode before and after boronization and in the scenario with fast current ramp-down in ohmic H-mode. A maximum value of of 2.0% and of 2.0 were achieved. The limit achieved reveals itself as a `soft' (non-disruptive) limit. The stored energy slowly decays after the current ramp-down. No correlation was found between beta restriction and MHD phenomena. Internal transport barrier (ITB) formation was observed in ohmic H-mode. An enhancement factor of 2.0 over ITER93H(ELM-free) was found in the ohmic H-mode with ITB.

Study of toroidal current penetration during current ramp in JIPP T-IIU with fast response Zeeman polarimeter

Toroidal current penetration is studied in current ramp experiments on theJIPP T-IIU tokamak. The poloidal magnetic field profile in the peripheral region of a plasma (0.5 ⩽ r/a ⩽ 1.0) has been measured directly with a newly developed fast response Zeeman polarimeter. The experimental results indicate that a skin effect of the toroidal current density is clearly observed during both the current ramp-up (CRU) and the current ramp-down (CRD) experiments. The experimentally obtained toroidal current density profiles are well described by the profiles calculated on the assumption of neoclassical electrical resistivity, although the difference between the neoclassical electrical resistivity and the Spitzer electrical resistivity is not clearly distinguished because of the relatively collisional peripheral plasma region. Quasi-linear Δ' analysis of tearing modes for the measured current density profile is consistent with the time behaviour of the coherent magnetohydrodynamic (MHD) modes such as m = 4/n = 1 or m = 3/n = 1 (m is the poloidal mode number and n is the toroidal mode number) often observed during the CRU phase. An obvious enhancement of the current penetration by the MHD modes is not detected in this ramp-up experiment, where the relative amplitude of the poloidal field fluctuations is less than 0.8%.

Current profile measurements with motional Stark effect polarimeter in the JT-60U tokamak

A motional Stark effect polarimeter with five viewing points has been installed to obtain current profiles in the JT-60U tokamak. Offset angles caused by Faraday rotation were calibrated by rapidly moving a plasma, keeping the q profile unchanged. Polarized background light was observed and its effect was corrected by detecting the light which had a wavelength close to that of the beam emission. Current profiles in a current rampdown discharge were measured, and the q value at the inversion radius of the sawtooth oscillation agreed with the theoretical value (unity) within 10%.

Review of D-T Results from TFTR

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date, the maximum fusion power has been 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-{beta}{sub p} discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MWm{sup {minus}3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITTER). The energy confinement time, {tau}{sub E}, is observed to increase in D-T, relative to D plasmas, by 20% and the n{sub i}(O){center_dot}{tau}{sub E} product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-{beta}{sub p} discharges. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations assuming classical confinement. Measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from helium gas puffing experiments. The loss of energetic alpha particles to a detector at the bottom ofmore » the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. ICRF heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. D-T experiments on TFTR will continue both to explore the physics underlying the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.« less

Disruption Avoidance on TFTR

Disruptions on the Tokamak Fusion Test Reactor (TFTR), especially those occurring at high stored energy, result in lost experimental run time because many discharges are required to regain wall conditions necessary for good plasma performance. A variety of disruption types have been observed on TFTR. These include density-limit disruptions, those caused by a high influx of impurities, those occurring during the current rampdown, those resulting from locked modes, and those occurring at high normalized β (β N = β T αβ T /I p ). A combination of operational experience and limiter development has helped to avoid many potential disruptions. However, the experimental goal of high fusion power production engenders the risk of high-β N disruptions. A system to limit β N by reducing the neutral beam power as a preprogrammed β N limit is reached is now in use to help avoid high-β N disruptions. Operational issues of disruption avoidance, the β N feedback system, the limitations and possible improvements of the system are discussed.

Overview of DT results from TFTR

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date (September 1995), the maximum fusion power has been 10.7 MW, using 39.5 MW of neutral beam heating, in a supershot discharge and 6.7 MW in a high beta P discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MW/m3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER). The energy confinement time tau E is observed to increase in DT, relative to D plasmas, by 20% and the n1(0).T1(0). tau E product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter H mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high beta P discharges. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations assuming classical confinement. Measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from helium gas puffing experiments. The loss of energetic alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha particle driven instabilities has yet been observed. ICRF heating of a DT plasma, using the second harmonic of tritium, has been demonstrated. DT experiments on TFTR will continue both to explore the physics underlying the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor

Emergency discharge quench or rampdown by a noble gas pellet

Fast discharge quench in ITER using injection of a krypton pellet is considered. A zero dimensional analysis predicts a hundred milliseconds quench time for a 10 mm pellet injected with a velocity of 3 km/s. No significant runaway electrons are expected in the Parail-Pogutse approach improved by the 'avalanche effect'. Temperature shock due to radiation onto the beryllium first wall is not greater than 850 K for 10 ms close to the start of the current rampdown. This seems acceptable for the wall performance

Review of deuterium–tritium results from the Tokamak Fusion Test Reactor

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.

Experiments on the current rampdown phase in the STOR-M tokamak for AC operation

The current rampdown phase followed by non-disruptive current termination is studied for future alternating current (AC) operations in the STOR-M tokamak. The plasma current is smoothly reduced to zero without disruptions with the rampdown rate of around 1.5 kA ms-1. The plasma density in the current rampdown phase is controlled by gas puffing. The plasma current decreases, maintaining the relationship 1/qa approximately 0.01 nR/Bt (10-18 m2T). At current termination, Ip=0, a residual plasma density (0.58-2.9)*1018 m-3 is observed with a corresponding Murakami parameter nR/Bt=(0.41-2.1)*1018 m-2 T-1. After current termination, the low-density plasma remains for 5 to 15 ms, which would allow AC operation in the STOR-M tokamak if the radiation barrier does not affect the subsequent current evolution.

Plasma processes accompanying fast current decrease in the Tuman-3 tokamak

The authors report an investigation of the influence of radial current distribution on the Tuman-3 tokamak plasma parameters under the conditions of ohmic heating with a twofold current decrease for a period 4-6 times shorter than the magnetic field diffusion time-scale. A slight change in the parameters was observed in the current ramp-down phase. The characteristic relaxation time of the plasma parameters after the current drop was considerably larger than the energy confinement time and comparable with the field diffusion time-scale. The influence of the current ramp-down on the boundary plasma is also studied. Narrowing of the current density profile is accompanied by a 1.5-2-fold rise of the energy confinement time compared with its steady-state value.

A study of the Tokamak de Varennes plasma during fast current ramp-down: Experiment and simulation results with the TSC code

Simulation results obtained with the axisymmetric Princeton Tokamak Simulation Code TSC are compared with experimental results for the time evolution of the Tokamak de Varennes plasma. The code correctly predicts experimental conditions; of special interest are the favourable comparisons of the simulation with experiments during fast current ramp-down (from 250 kA in 40 ms, three times faster than the L/R time).

Erosion of long-term wall samples in JET

The erosion of carbon and the deposition of carbon and impurities have been determined for graphite samples exposed at the wall of JET during the six months of operation in 1985. Graphite erosion was determined using SIMS analysis of the depth change of implanted 13C marker layers in the samples. Impurity deposition was measured using SIMS and Rutherford backscattering. The samples showed high erosion ranging from 0.2 to 0.6 μm at various locations, with significant poloidal variation in both erosion and deposition. The deposition of substantial quantities of metal impurities and carbon was related to a probe accident and a deliberate heavy carbonization procedure during the operation period. The highest observed erosion exceeds the expected sputter erosion due to charge exchange neutrals by more than an order of magnitude, unless neutral fluxes are enhanced at the inner wall due to ions arriving along open field lines during current ramp-down.