G. Maddaluno

FTU results with a liquid lithium limiter

Since the end of 2005 most of the plasma?wall interaction experiments on FTU have been focused on the possible use of liquid lithium as the plasma facing material. Liquid lithium limiter is an active method to deposit, during the plasma discharge, a lithium film on the walls with prolonged beneficial effects. Reliable operation with very clean plasmas, very low wall particle recycling, spontaneous peaking of the density profile for line-averaged density values 1.0times 10^{20},{rm m}^{-3} SRC=http://ej.iop.org/images/0029-5515/51/7/073006/nf381122in001.gif/> have been obtained. These results have allowed us to extend the density limit to the highest value so far obtained ( at Ip = 0.7?MA and BT = 7.1?T, qa = 5.0, by gas puffing only) and to increase the energy confinement time by almost 50% with respect to the average value of 50?ms of the old ohmic FTU database. An accurate analysis of these plasmas has been carried out by means of a gyrokinetic code to establish the role of collisionality and density gradients on the observed phenomenology.

FAST plasma scenarios and equilibrium configurations

In this paper we present the fusion advanced studies torus (FAST) plasma scenarios and equilibrium configurations, designed to reproduce the ITER ones (with scaled plasma current) and suitable to fulfil plasma conditions for integrated studies of plasma‐wall interaction, burning plasma physics, ITER relevant operation problems and steady state scenarios. The attention is focused on FAST flexibility in terms of both performance and physics that can be investigated: operations are foreseen in a wide range of parameters from high performance H-mode (toroidal field, BT, up to 8.5 T; plasma current, IP, up to 8 MA) to advanced tokamak (AT) operation (IP = 3 MA) as well as full non-inductive current scenario (IP = 2 MA). The coupled heating power is provided with 30 MW delivered by an ion cyclotron resonance heating system (30‐90 MHz), 6 MW by a lower hybrid system (3.7 or 5 GHz) for the long pulse AT scenario, 4 MW by an electron cyclotron resonant heating system (170 GHz − BT = 6 T) for MHD and localized electron heating control and, eventually, with 10 MW by a negative neutral ion beam (NNBI), which the ports are designed to accommodate. In the reference H-mode scenario FAST preserves (with respect to ITER) fast ion induced as well as turbulence fluctuation spectra, thus addressing the cross-scale couplings issue of micro- to meso-scale physics. The non-inductive scenario at IP = 2 MA is obtained with 60‐70% of bootstrap current and the remaining by LHCD. Predictive simulations of the H-mode scenarios have been performed by means of the JETTO code, using a semi-empirical mixed Bohm/gyro-Bohm transport model. Plasma position and shape control studies are also presented for the reference scenario.

Edge plasma physics issues for the Fusion Advanced Studies Torus (FAST) in reactor relevant conditions

To have reliable predictions of the thermal loads on the divertor plates and of the core plasma purity in the proposed Fusion Advanced Studies Torus (FAST) tokamak, numerical self-consistent simulations have been made for the H-mode and steady-state scenario by using the 2D multi-fluid code COREDIV. In the simulations full W plasma facing components, foreseen for basic operation, as well as liquid lithium divertor targets have been considered. Impurity seeding, for reducing divertor heat loads, was allowed. The overall picture shows that, marginally in the intermediate and, necessarily in the high density H-mode scenarios (average density ne = 2 ? 1020?m?3 and 5 ? 1020?m?3, respectively), impurity seeding should be foreseen with W as target material; however, only a small amount of Ar (0.03% atomic concentration), not affecting the core purity, is sufficient to maintain the divertor peak loads below 18?MW?m?2, which represents the safety limit for the W monoblock technology, presently accepted for the ITER divertor tiles. Li always needs additional impurities for decreasing divertor heat loads. At low plasma densities (but ?1.3 ? 1020?m?3), typical of steady-state regimes, W alone is effective in dissipating the input power by radiative losses, without excessive core contamination. Impurity seeding would lead to excessive W sputtering by Ar and too high Zeff. The impact of unmitigated giant (1.5?MJ) type I edge localized modes on the W divertor targets was also analysed: the resulting maximum energy load of 1?MJ?m?2, larger than the tolerable one by a factor of 3, seems not difficult to recover by foreseen mitigation tools.

Optimization of the FTU toroidal limiter shape

The FTU machine is now operating with an Inconel/stainless steel poloidal limiter. The necessity to withstand a large heat flux in the radio-frequency heating operational phase and to minimize plasma edge asymmetries has induced the feasibility of a toroidal limiter to be assessed. The take into account heat loads due to inevitable misalignments in the radial positioning of the limiter sectors, a 3-D analysis of the problem was carried out. Within this analysis, the shadowing between adjacent sectors has been found to be the most critical factor because of the shortness of energy decay length. In spite of the 3-D optimization of the limiter surface, peak heat load strongly increases with the safety factor at the edge and reaches up to a factor of 4 with respect to the average heat load.

Effects of wall boron coating on FTU plasma operations

To achieve good performance on FTU in a large density range (0.3-6.0 × 10 20 m -3 ), boronization with a mixture of He (90%) and B 2 H 6 (10%) (diborane) as the feeding gas has been tested with thermal loads on the limiter surface up to 2.5 MW/m 2 . With boronized limiter (TZM alloy with 98% of Mo) and walls (SS AISI 304), the total radiated power drastically drops from 70-90% down to 35-45% and the Z cn decreases from 6.0 to 2.2 at 0.3-0.4 × 10 20 m -3 related to a strong reduction of heavy-metal concentration and to the getter effect of boron on oxygen (<0.5%). During this phase the action of the boron film as particle reservoir and its quick saturation due to the low temperature of FTU walls makes it difficult to obtain reproducible plasmas. Another consequence of boronization is the large dilution of the plasma with the hydrogen particles released from the B film. All these effects decrease after about 60 discharges when boron is eroded by the limiter but it is still present on the chamber walls. During this phase which lasts for more than 500 discharges, oxygen concentration does not increase at all and metal influx is lower than before boronization because the physical sputtering by oxygen ions and atoms is strongly reduced.

Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components ☆

Abstract The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime.

Activation and damage by runaway electrons on FT limiter

Abstract Activation and metallographic analyses have been used in order to characterize the runaway electron interaction with the stainless steel limiter of the FT machine. Electrons with energy up to 20 MeV are found to hit the limiter with an average radial penetration ranging from 0.14 to 0.33 cm. For an incidence time of 50 ms a power load ≧ 12 kW/cm2 is deduced from the depth of the molten layer.

Heat fluxes and energy balance in FT limiters

Abstract The main poloidal limiter used in FT during 1983 consists of 38 mushroom shaped pieces tightened onto a supporting structure. All items are made of stainless steel (AISI 316). Measurement of weight losses and also observation of damage have been carried out. An e-folding length of λ E λ E cm was estimated. A global energy balance gives a maximum total power on the limiter of about 300 kW. Weight loss measurements on the fraction of the limiter involved in disruption reveal that only a fraction of the thermal energy of the plasma is absorbed by the material suggesting the presence of some self-protecting mechanisms around the limiter.

Runaway-limiter interaction in the FTU tokamak during disruptions

Abstract Photoneutron production during disruptions was monitored on the FTU machine by a set of calibrated BF 3 proportional counters. Both hydrogen and deuterium discharges covering a wide density range were examined. The functional dependence of the photoneutron production on n e , Z eff and I p is in agreement with the Dreicer generation, being the main source of runaway electrons during disruptions in FTU. No detectable photoneutron production was observed at a plasma density higher than 1.4xa0×xa010 20 m −3 , corresponding to Z eff values close to 1. The typical runaway energy was evaluated from the curve of the photoneutron yield per electron vs. energy for disruptions characterized by a hesitation in the plasma current decay. Damage by disruption generated runaways was found to be concentrated around the inboard equatorial plane as suggested by temperature and activation measurements.

Energy deposition on the FTU poloidal limiter during disruptions

Abstract We present the first results of the program for the characterization of the thermal flux on the FTU poloidal limiter during disruptions. Data on power fluxes are obtained by using an infrared detector and a set of thermocouples. Two peaks in the limiter thermal load, corresponding to the thermal( up to 500 MW/m 2 ) and magnetic quenches, are well resolved by the infrared detector allowing the time correlation with other fast diagnostic measurements. The dependence on the main plasma parameters of the intensity and time evolution of the thermal flux to the limiter is discussed.