Gas Box Research Articles

Tungsten filament mock-ups for gas box liner

Carbon impurities create carbon films saturated by tritium and deuterium onto the gas box liner. The reference gas box liner formed by chevron-shaped tungsten tiles is capable of depositing and/or recombining only 5% of the carbon impurities. The innovative solution presented here increases within the range 10 2 to 10 3 the surface interacting with the pumped gas by using an array of filaments, either disordered (felt) or ordered (spiral). This paper describes the experiments carried out to prepare felt/spiral samples and indicates possible ways of fabricating a representative gas box liner (GBL) mock-up.

Design of a radiative semi-transparent liner for the ITER divertor cassette

The increased safety concern, resulting from the accumulation of tritium in the redeposited carbon-dominated films forming on the surfaces of the ITER divertor private region plasma facing components, has stimulated an important change of the gas box liner design. A significant modification of the ITER divertor cassette design was undertaken in the summer of 1997 with the removal of the so-called ‘wings’. This highlighted the necessity to protect the cold (150°C) pumping duct from radiated heat and particle flux from the plasma chamber. A cold duct has the potential to trap unacceptably high levels of tritium in the redeposited carbon films produced from erosion of the vertical target. To provide protection that minimises the uptake of tritium, a new semi-transparent liner operating with hot tungsten (W) tiles was proposed by the ITER Joint Central Team (JCT) and was developed by the RF Home Team. This liner design promises to minimise the amount of tritium trapped in it by operating with the surface temperature of the W in the range 800–1200°C. Radiatively cooled tiles on the surface of the liner maintain the surface temperature between these limits, for a surface heat flux of between 0.1 and 0.35 MW/m 2, a range which takes into account the uncertainties in the predicted operational heat loads. This paper reports on the predicted operational regime of the liner which allows effective transformation of incoming atomic tritium and carbon into stable hydrocarbons, that can be pumped safely to the reprocessing plant. The role of hydrocarbons (as molecules and radicals) is assessed. A sensible parametric analysis of tritium retention, on the liner surface and on cold areas behind the liner, was performed for different conditions and operational parameters.

Novel designs of the ITER gas box liner aimed at reducing tritium codeposition

Carbon impurities create carbon films saturated by tritium and deuterium onto the gas box liner. The reference gas box liner (GBL) formed by chevron shaped tungsten tiles is capable of depositing and/or recombining only 5% of the carbon impurities. The innovative solution presented here increases within the range 10 2–10 3 the surface interacting with the pumped gas by using an array of filaments, either disordered (felt) or ordered (spiral). This paper describes the experiments carried out to prepare felt/spiral samples and indicates possible ways of fabricating a representative gas box liner mock-up.

Towards steady state tokamak operation with double transport barriers

Core confinement and the overall performance of the tokamak have been considerably enhanced with the combination of two improved confinement regimes. Internal transport barriers characteristic of the optimized shear regime and an edge transport barrier of the high confinement H mode regime have been superposed in the double barrier (DB) mode. In DT discharges the DB mode has resulted in a fusion gain Q a factor of 2 higher than in conventional sawtoothing steady state ELMy H mode plasmas. The DBmode has been routinely established in the new Gas Box divertor configuration on JET. Off-axis lower hybrid current drive and edge radiation cooling through impurity seeding have been used for current and pressure profile control. This has made it possible to increase the performance up to βN = 2.4 and H89 = 2.7 and to extend flat-top phases up to a duration of three energy confinement times, approaching steady state conditions. Modelling studies of DB mode DT operation on JET predict a fusion power in the range 20-30 MW. In ITER transport code modelling, steady state operation with a fusion power output of 1.3 GW is obtained for advanced tokamak scenarios in the DB mode.

High performance with modified magnetic shear in JET DD and DT plasmas

Internal transport barriers (ITBs) in which both the ion thermal diffusivity and electron thermal diffusivity are substantially reduced have been observed in JET. Such discharges have been obtained with DD and DT plasmas. Central ion temperatures of 40 keV and plasma pressure gradients of 106Pa/m were observed in DT plasmas leading to a fusion triple product ni0Ti0τE = 1.1 × 1021 m-3·keV·s and producing 8.2 MW of fusion power. ITBs have been produced in both the MkII and the new Gas Box divertor configuration with similar behaviour. With the Gas Box divertor an L mode edge has so far only been produced using edge radiation cooling. For the first time,ITBs have been triggered by radiating about 40% of the power with a krypton puff. A possible scaling of the power needed to trigger an ITB with magnetic field is suggested.

A report on the first remote handling operations at JET

The first fully remote handling operations on a fusion Tokamak have been completed at JET. The JET MkIIa divertor has been successfully removed and replaced with the MkII gas box divertor configuration using only remote handling methods inside the JET torus. This shutdown required the handling of over 450 components over a 15 week period operating 20 h per day and 6 days per week. The JET remote handling equipment functioned and performed almost flawlessly with only two significant faults. The total operational time lost due to equipment problems was less than 3 days. The operations proceeded safely and without significant problems. No time was lost due to either operator induced equipment problems or operational errors. A number of new tasks were successfully included in the operations at short notice. Some new tasks were identified immediately before the shutdown but with sufficient time available to test their feasibility using the mock-up facility. Others were only evident after the shutdown had started. The successful implementation of the shutdown and in particular of the unprepared tasks has been due to the high degree of adaptability provided by the JET man-in-the-loop remote handling approach.

Physics of high performance JET plasmas in DT

The Joint European Torus (JET) has recently operated with deuterium-tritium (DT) mixtures, carried out an International Thermonuclear Experimental Reactor (ITER) physics campaign in hydrogen, deuterium, DT and tritium, installed the Mark IIGB `Gas Box' divertor fully by remote handling and started physics experiments with this more closed divertor. The DT experiments set records for fusion power (16.1 MW), ratio of fusion power to plasma input power (0.62, and 0.95 ± 0.17 if a similar plasma could be obtained in steady state) and fusion duration (4 MW for 4 s). A large scale tritium supply and processing plant, the first of its kind, allowed the repeated use of the 20 g of tritium on-site to supply 99.3 g of tritium to the machine. The H mode threshold power is significantly lower in DT, but the global energy confinement time is practically unchanged (no isotope effect). Dimensionless scaling `wind tunnel' experiments in DT extrapolate to ignition with ITER parameters. The scaling is close to gyro-Bohm, but the mass dependence is not correct. Separating the thermal plasma energy into core and pedestal contributions could resolve this discrepancy (leading to proper gyro-Bohm scaling for the core) and also account for confinement degradation at high density and at high radiated power. Several radiofrequency heating schemes have been tested successfully in DT, showing good agreement with calculations. Alphaparticle heating has been clearly observed and is consistent with classical expectations. Internal transport barriers have been established in optimized magnetic shear discharges in DT and steady state conditions have been approached with simultaneous internal and edge transport barriers. First results with the newly installed Mark IIGB divertor show that the in-out symmetry of the divertor plasma can be modified using differential gas fuelling, that optimized shear discharges can be produced and that krypton gas puffing is effective in restoring L mode edgeconditions and establishing an internal transport barrier in such discharges.

The Remote Exchange of the Jet Divertor

In 1997 a series of experiments were performed in the JET machine using deuterium-tritium (D-T) mixtures and resulting in discharges with record breaking fusion power and fusion energy. The experiments demonstrated a key technology required for fusion, namely the on-line operation of a tritium fuel re-processing plant.These experiments left the inside of the JET vessel inaccessible to manned access for approximately one year. During this time, the complete Mark IIA divertor, a major system within the torus, was successfully removed and replaced with a new divertor design, the Mark II Gas Box divertor, using only remote handling techniques. This was the first application of the JET remote handling system and a demonstration of a further key ITER technology.The paper explains the methodology and operational approach taken to achieve the results using the remote handling system developed at JET. It describes the remote handling equipment including the force-reflecting servomanipulator, the specialised tools designed, the facilities needed, and the trials, planning and training carried out to ensure the safe, reliable and rapid completion of the remote handling tasks. The planned tasks are outlined including the execution of the novel procedure for a remote, sub-millimetre precision, dimensional survey of the divertor support structure using digital photogrammetry. Furthermore the paper shows how the adaptability of the system was used to successfully undertake a large number of unplanned tasks including the removal of damaged tiles, a damaged diagnostic system and the vacuum cleaning of diagnostic windows.

Detailed conceptual design of the fusion experimental breeder

A detailed conceptual design for a Fusion Experimental Breeder (FEB) was constructed. It was He-cooled with liquid lithium/Li 2O as a tritium breeder and the machine parameters were as follows: a major radius of 4 m, a minor radius of 1 m, magnetic field on axis 5 T, a plasma current 5.7 MA, and a neutron wall loading of 0.42 MW m −2. Various aspects of the core plasma were studied comprehensively. CAD software was used to create a complete set of drawings, including 3D-perspective ones. Emphasis was given to the design of the nuclear components. Two versions of the blanket were designed and attention was paid to the design of the blanket structure. Detailed analyses were performed, i.e. 3D neutronics analyses with the latest ENDL, 3D thermo-mechanic analyses, probability risk assessment of the main cooling system failure, 3D shield analyses, etc. A gas box divertor is at present being studied. Radioactivity related parameters were calculated, including the nuclear waste disposal rating of the first wall and inventories of 232U and 237NP, etc. The cost of FEB was estimated with the calculation of the ITER capital cost as a reference. Based on this design, an engineering outline was started in 1996.

Engineering experience in JET operations

The inherent flexibility of JET's original concept has permitted several engineering upgradings and modifications, to address a large variety of plasma and fusion physics issues. The most recent major modification has been the installation of an axisymmetric single-null pumped divertor (Mark I), successfully operated in the experimental period 1994–1995. Following the divertor optimization programme a new, more closed, divertor configuration has now been installed (Mark II), which has shown a better power handling capability and substantially improved neutral particle retention. A key feature of the new design is the possibility to replace the divertor target plate structure using full remote handling techniques following extended D-T operations. Toroidal asymmetries of vessel forces due to Vertical Displacement Events (VDE) and halo currents were experienced since 1994, leading in some cases to sideways movements of the vessel of 7 mm. This has required modification and upgrading of the vacuum vessel support system. Gap control of plasma position and shape and machine protection systems have been developed further, leading to increased experimental availability. Future development foresees the installation of a Mark II Gas Box divertor structure, while studies are underway to increase the toroidal field capability from 3.45 to 4 T and the additional heating power by increasing the NB injector output from 80 kV, 60 A to 120 kV, 60 A and by using wide band matching for ICRF.

Plasma-neutral gas interaction in a tokamak divertor: Effects of hydrogen molecules and plasma recombination

We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 × 10 −10 cm 3/ s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a ‘gas box’ divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes.

Plasma–neutral interaction in tokamak divertor for ‘‘gas box’’ neutral model

Plasma flow through the gas cloud in a tokamak divertor for ‘‘gas box’’ divertor geometry and Knudsen regime of neutral transport is investigated. It is shown that similar to the neutral models that have considered previously, (i) plasma parameters near the target is sensitive to the energy flux into the hydrogen recycling region and can change rapidly, resulting in bifurcation-like behavior, which might be interpreted as a transition to detached regime, (ii) plasma flux onto the target starts to decrease at a very low plasma temperature near the target, while a strong pressure drop already occurs. At low plasma temperature near the target the recombination processes can significantly alter the plasma flux onto the target.

The TMX heavy ion beam probe

A heavy ion beam probe has been used to measure the radial space potential distribution in the central cell of TMX. This was the first beam probe system to utilize computer control, CAMAC instrumentation, and fast time response for broadband fluctuation capabilities. The fast time response was obtained using off-line processing of the energy analyzer detector signals and wideband transimpedance amplifiers. The on-axis space potential was found to be 300-400 V, with /spl phi//sub e//T/sub ec//spl sim/8. The radial potential profile is parabolic when gas box fueling is used. The frequency of observed fluctuations was found to agree with the E/spl times/B plasma rotation frequency during the discharge. The measured Tl/sup ++/ secondary ion current level is consistent with calculations, given reasonable assumptions for beam attenuation. >

In situ analyzer calibration methods for heavy ion beam probes installed on stellarator-like devices

An absolute calibration of an installed heavy ion beam probe (HIBP) energy analyzer is possible by detecting secondary ions that come from a source at a known location that is maintained at a known potential. It is also necessary that the magnetic field that exists during the calibration be the same magnetic field that exists when a plasma is present. These conditions can be met on stellarators, heliotrons, and torsatrons, which produce their magnetic configurations with external coil sets. Since no internal plasma current is required, suitable sources for producing secondary ions by interaction with the primary beam can be placed inside the vacuum vessels of these devices at known locations. Secondary ions can be produced by the interaction of the primary beam with thin films, gas box probes, electron beams, or neutral gas filling the vacuum vessel. Details of a spherical mesh probe that combines the advantages of several of these methods are given.

Laboratory simulation of atmospheric corrosion by SO2—I. Apparatus, electrochemical techniques, example results

The atmospheric corrosion by SO2 has been simulated in the laboratory using a pollution gas box, PGB, which is described. Both two-electrode galvanic cell type atmospheric corrosion monitors, ACMs (Zn/Fe, cast Zn-55% Al/Fe, rolled Zn-55% Al/Fe), and three-electrode electrochemical cell type ACMs (Zn, Fe, cast Zn-55% Al, rolled Zn-55% Al) have been constructed using techniques modified from those available in the literature. Techniques for carrying out PGB experiments on metal samples in flowing laboratory air containing 0–1 ppm [SO2] are described. Techniques for weighed sample mass loss and ACM electrochemical tests (galvanic current, impedance, corrosion current) are given along with examples of typical electrochemical results. Methods for calculating mass loss from these electrochemical tests are indicated.

A segmented high pressure gas target for (n, p) studies at TRIUMF

We have developed an (n,p) target chamber consisting of four target layers sandwich between five detector layers. The two] central targets are gas cells separated from the detector gas volume by thin windows. The two outer targets are solids. The chamber has been used with isotopic targets of 20Ne, 3He and 15N at the TRIUMF (n, p) facility. With the gas cells and gas box operated at 20 atm (20.3 bar) substantial target thickness can be achieved, while the contamination of (n, p) spectra from extraneous materials is minimized. The simultaneous accumulation of data from the two gas cells and two solid targets has the advantages of: 1) normalization by including targets of known cross section in the target stack and/or 2) obtaining the spectral shape of the incident neutron flux by choosing C and CH 2 for the solid targets.

Radio-frequency wave interchange stability experiments below the ion cyclotron frequency

It has been demonstrated in the Phaedrus-B tandem mirror [Phys. Fluids 31, 714 (1988)] that radio-frequency (rf) waves applied below the ion cyclotron frequency (ω/Ωi ≊0.75) can stabilize a mirror plasma against the interchange instability. Results presented here demonstrate stabilization over a range of frequencies (0.65≤ω/Ωi≤0.8) with the mMHD =−1 instability stabilized and higher-order (‖mMHD‖≥2) low-amplitude modes present. Radial profiles of the rf magnetic fields at a number of axial locations as well as parallel wavenumber measurements indicate mrf =−1 slow shear Alfvén wave excitation over this frequency range. Measurements at frequencies below 0.75Ωi indicate attenuation of rf waves by gas baffles and a gas box.

Simulation of ion confinement during ion cyclotron heating in the TMX-U tandem mirror

Particle and energy confinement of ions during ion cyclotron heating (ICH) in the centre cell of a tandem mirror is simulated by a Monte Carlo code. The interaction of the ion cyclotron wave with the ions is calculated using reduced equations of motion, and Coulomb collisions are modelled using a Monte Carlo algorithm. An axial electrostatic potential profile is generated by an iteration procedure, assuming Maxwellian electrons. The axial profile of neutral gas which fuels the ions from a localized gas box is determined from a neutral particle Monte Carlo code. Steady state calculations are reported for the Tandem Mirror Experiment Upgrade (TMX-U). The calculated energy confinement time τE is about twice that measured for off-midplane fuelling. The calculated τE can be decreased to the experimental value by either introducing extra gas at the ICH antenna to model the effect of ions striking the antenna or assuming anomalous radial loss. However, the enhanced radial loss also increases the end loss current well above that measured. Experimental and calculated τE agree more closely for midplane fuelling with heating from one antenna.

Plasma production with rotating ion cyclotron waves excited by Nagoya Type-III antennas in RFC-XX

A method of plasma production by ion cyclotron wave heating has been developed in RFC-XX. Nagoya Type-III antennas were used for wave excitation, and gas was supplied through a gas box. The effect of the rotating field excitation on plasma production and heating was investigated. In the m = − 1 rotational mode (rotation in the direction of ion cyclotron motion), the plasma density profile is flat within the gas box bore with a line integrated density nℓ = 3 × 1013cm−2, and the ion temperature is Ti ≈ 150 eV. For the m = +1 mode, a high density plasma was obtained with a different profile having a peak at the centre. In this mode, the line integrated density is nℓ = 3 × 1014cm−2, with the peak density n(0) = 7 × 1013cm−3.

Gas injection system in the Tara central cell

The gas injection system in the Tara central cell has been redesigned to overcome problems associated with fueling in the high field region at the ends of the central cell. Previous experimental studies indicated that plasma ions created in the region between the gas injection point and an ICRH antenna were lost radially before passing to the downstream side of the antenna. In the new system gas is injected at two diametrically opposite ports into a cylinder (‘‘gas box’’) located near the midplane, between the two double half turn antennas. A maximum in the magnetic field at the midplane divides the central cell into two R=2 mirrors. Ions trapped in one of these mirrors must scatter into the loss cone before passing into the fueling region, where the loss rate due to charge exchange is highest. This configuration produces separate regions for ionization (at the gas box) and for heating (near the magnetic field minimum). Pumping of neutral gas by the central cell plasma reduces the amount of neutral gas flux into the axicell to acceptable levels. Gas utilization efficiencies are roughly twice as high as in the original configuration.