N. Bekris

Erosion/deposition issues at JET

Deposition and H-isotope retention in JET is highly asymmetric, with deposition predominantly in the inner divertor, where flaking deposits form on water-cooled louvres shadowed from the plasma. The asymmetry implies drift in the SOL of JET from outboard to inboard under normal operating conditions, which has been measured. In order to model the amounts of deposition, assumptions have to be made about the transport at the inner target. Analysis of divertor tiles shows that material from the main chamber travels along the SOL to the inner divertor wall, from where carbon is preferentially removed leaving a beryllium-rich film. The carbon travels to shadowed areas (such as the louvres) where deposits with high H-isotope content accrue. The analysis indicates that chemical processes must be important.

Tritium depth profiles in graphite and carbon fibre composite material exposed to tokamak plasmas

Abstract Tritium inventories in the plasma-facing surface and the bulk of tiles were investigated with a highly sensitive, accurate full-combustion technique and a PIN-diode method. Examined were (i) a tokamak fusion test reactor (TFTR) graphite tile (D–D plasmas), (ii) a JET graphite tile (low-tritium D–T plasmas), and (iii) several JET carbon fibre composite (CFC) divertor tiles as well as a graphite limiter tile (all high-tritium D–T plasmas). Whilst the bulk tritium concentration in graphite tiles appears to remain at very low levels (about 0.3% of the total tritium) the tritium bulk concentrations in CFC divertor tiles can be as high as three times that in the surface layer. The latter is attributed to plasma-induced trapping of tritium between the fibre planes of CFC in the hot divertor zone. In addition to carbon/hydrogen co-deposition, this contribution constitutes another important source of tritium inventory in the torus that so far had not been recognised.

Tritium retention of plasma facing components in tokamaks

Abstract The areal distribution of tritium retention in tiles from TEXTOR, TFTR, JT-60U and JET has been measured via the imaging plate technique and the results are discussed from the perspective of carbon–hydrogen chemistry. It is found that the observed tritium distribution clearly shows asymmetries in poloidal and toroidal directions and also reflects the local temperature history of the analyzed tiles. We show the first clear evidence of the loss of high energy tritons by toroidal magnetic field ripple. We distinguish three different contributions to tritium retention in tokamaks with carbon plasma facing components: high energy tritons escaping from the core plasma, low energy ions and neutrals from the edge plasma, and molecular tritium from gas fueling. These components are retained at different depths and with different concentrations. Tritium from the edge plasma dominates the retained inventory but could be reduced if the surface temperature was higher. We propose tokamak operation with plasma facing components above 1000 K as a possible way to reduce the tritium inventory.

Tritium profiles in tiles from the first wall of fusion machines and techniques for their detritiation

Abstract Tritium profiles on a TFTR graphite tile exposed to D–D plasmas and a JET graphite tile from the first tritium campaigns were examined by full combustion, thermogravimetry and thermal desorption. Combustion measurements revealed that >98.9% of the tritium is trapped in a layer 95% under a stream of moist air at about 400°C. A large fraction of tritium can be removed from the tile surface with adhesive tape.

Tritium Removal from JET and TFTR Tiles by a Scanning Laser

Fast and efficient tritium removal is needed for future DT machines with carbon plasma facing components. A novel method for tritium release has been demonstrated on codeposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave Nd laser beam was focused to ≈100 W/mm2 and scanned at high speed over the codeposits, heating them to temperatures ≈2000 °C for about 10 ms in either air or argon atmospheres. Up to 87% of the codeposited tritium was thermally desorbed from the JET and TFTR samples. Fiber optic coupling between the laser and scanner was implemented. This technique appears to be a promising in situ method for tritium removal in a next step DT device as it does not rely on oxidation, and avoids the associated deconditioning of the plasma facing surfaces and the expense of processing large quantities of tritium oxide.

Beryllium and carbon films in JET following D-T operation

After the D-T operation (DTE-1 campaign) at JET a large number of limiter and divertor tiles were dismounted from the torus for ex situ examination. The relative distributions of deuterium, tritium, beryllium and carbon are presented and discussed. Significant asymmetry observed in the distribution of erosion and deposition zones indicates preferential flow of the deuterium background plasma and impurities towards the inner divertor leg. The comparison of the beryllium content on the limiter tiles from the main chamber and the content of this element on the inner divertor tiles clearly proves the beryllium erosion from the main chamber wall and its transport to the divertor. However, no beryllium is detected in the shadowed regions of the divertor where the formation of thick and fuel-rich carbon films occurs. This is interpreted in terms of different mechanisms governing the erosion and transport of Be and C. The results allow a conclusion that the operation with a full beryllium wall would lead to a significantly decreased fuel inventory due to removal of the carbon source.

Tritium in plasma facing components

Recent results on measurements of tritium and other hydrogen isotopes in first wall materials of large tokamaks are discussed and evaluated. Data on the in situ and ex situ release of tritium from plasma facing components under different conditions are assessed.

Status and research progress at the Tritium Laboratory Karlsruhe

The Tritium Laboratory Karlsruhe (TLK) has been mainly designed for the conduction of technological experiments relevant to fusion under simulation of the conditions actually expected in the various tritium processing systems. Presently ongoing experimental and design work is almost exclusively oriented towards the needs of JET and ITER-H-FEAT. Basic research, while increasing in importance, plays predominantly a complementary role. This paper presents the most recent progress and developments related to the technological and applied experiments at the TLK.

Tritium depth profiles in 2D and 4D CFC tiles from JET and TFTR

Carbon fibre composite (CFC) is currently the candidate material for the vertical target tiles in the International Thermonuclear Experimental Reactor (ITER) divertor because of its superior thermomechanical properties. However, its affinity for hydrogen isotopes and their co-deposition with eroded carbon may severely limit ITER plasma operations. Recently tritium depth profiles in divertor tiles retrieved from the Joint European Torus have been obtained by the coring/full combustion technique. The results revealed that a surprisingly large fraction (up to 61%) of the retained tritium had diffused deep into the bulk of the tile, most probably between the woven sheets of the CFC. Additionally, the coring/full combustion technique has shown that only the surface tritium (few ten μm) is efficiently released by air baking while the bulk tritium is almost not affected. Baking the tile under air even at 500 °C does not detritiate the bulk.

Tritium distribution on the surface of plasma facing carbon tiles used in JET

Tritium surface profiles on divertor tiles used in JET were successfully determined applying imaging plate (IP) technique. The tritium intensities measured by IP were quite consistent with the previous tritium analysis made by full combustion measurements. Present results are summarized as follows. Most of the tritium in the divertor tiles was retained in co- or re-deposited layer and did not move because of their temperature was rather low. In addition tritium produced by D–D reaction are implanted in subsurface layers rather homogeneously, which seems the main tritium source for the inner divertor tiles with some thermal modification. Still there is another but very small tritium retention observed at the back side of the tiles, of which profile reflects the 2-D CFC structure, indicating preferential absorption and migration of tritium. 2003 Elsevier Science B.V. All rights reserved.