Only three tokamaks have operated with beryllium as the limiter or first-wall material. The first experi­ments were performed by UNITOR,45 which were then followed by ISX-B.46 Both tokamaks investi­gated the effects of small beryllium limiters on plasma behavior (UNITOR had side limiters at two toroidal locations and ISX-B had one top limiter) in support of the more ambitious beryllium experiment in JET (see below). The motivation to use beryllium came from the problem of controlling the plasma density and impurities when graphite was used.

Both UNITOR and ISX-B showed that once beryllium is evaporated from the limiter and coated over a large segment ofthe first wall, oxygen gettering leads to significant reduction of impurities. When the heat load on the beryllium limiter was increased to the point of evaporating beryllium, the oxygen concentration was decreased dramatically. Although the concentration of beryllium in the plasma was increased, its contribution to Zeff (the ion effective charge of the plasma Zeff provides a measure for impurity concentration) was more than compensated by the reduction of oxygen, carbon, and metal impu — rities.45 The plasma Zeff was observed to be reduced from 2.4 to near unity with beryllium. It must be noted that there was a negative aspect associated with beryl­lium operation during the ISX-B campaign. The reduction in plasma impurities was not observed until the limiter surface was partially melted causing beryllium to be evaporated and coated on the first wall. Once melting did occur, the droplets made subsequent evaporation more likely but hard to con­trol. The consequent strong reduction in plasma impurities associated with gettering then made dis­charge reproducibility hard to obtain. However, if a much larger plasma contact area is already covered with Be, one does not need to rely on limiter melting to obtain the beneficial effect ofberyllium. This effect could be achieved by using large area beryllium lim­iter, or coating the inside wall with beryllium which was the approach taken by JET when it introduced beryllium in 1989.

Large tokamak devices such as JET had found it very difficult to control the plasma density with graphite walls as the discharge pulse length got longer. Motivated by the frequent occurrence of a phenomenon that plagued the earlier campaigns — the so-called carbon blooms due to the overheating of poorly designed divertor tiles and the subsequent influx of carbon impurities in the plasma due to evaporation — JET decided to use beryllium as a plasma-facing material.

Thin evaporated beryllium layers on the graphite walls were used initially (~ 100 A average thickness per deposition) on the plasma-facing surface of the device. Subsequently, beryllium tiles were installed on the toroidal belt limiter.

The main experimental results with beryllium can be summarized as follows:

1. The concentration of carbon and oxygen in the plasma were 4-7% and 0.5-2%, respectively, when graphite was used as belt limiter. With a beryllium belt limiter, the carbon content was reduced to 0.5% and oxygen became negligible, because of oxygen gettering by beryllium. During ohmically heated discharges, the concentration of beryllium remained negligible even though beryllium was the dominant impurity.

2. While the value of Zeff was ^3 using the graphite limiter and auxiliary heating power of 10 MW, Zeff was ~1.5 even with additional heating powers of up to 30 MW with a beryllium limiter.

3. The fuel density control was greatly improved with the beryllium limiter and beryllium evapo­rated wall. Gas puffing to maintain a given plasma density was typically 10 times larger when using beryllium than graphite.

Following the beryllium limiter experience, diver­tor beryllium targets were installed in JET for two configurations. An extensive set of experiments with toroidally continuous X-point divertor plates was car­ried out inJET in the period 1990-1996 to characterize beryllium from the point of view of its thermomecha­nical performance, as well as its compatibility with

47-50

various plasma operation regimes.

In the JET Mk I experiments, melting of the beryllium tiles was reached by increasing (in a pro­gressive way) the power flux to a restricted area of the divertor target in fuelled, medium density ELMy H-mode discharges (Pinp ~ 12 MW). Large beryllium influxes were observed when the divertor target tem­perature reached 1300 °C. In these conditions, it became difficult to run low-density ELMy H-mode discharges (Pinp ~ 12 MW) without fast strike point movement (to achieve lower average power load) and the discharges either had very poor performance or were disrupted. However, no substantial plasma performance degradation was observed for medium density H-modes with fixed strike point position, or if fast strike point movement was applied in low — density H-modes, despite the large scale distortion of the target surface caused by the melt layer displace­ment and splashing due to the previous ^25 high power discharges48,51 (see Figure 2 52). This demon­strated that it was possible to use the damaged Be divertor target as the main power handling PFC if the

average power load was decreased, either by increas­ing plasma density and radiative losses, or by strike point sweeping. The damage did not prohibit subsequent plasma operation inJET, but would seri­ously limit the lifetime of Be PFCs in long-pulse ITER-like devices.

The latest results of the operation of JET with beryllium have been reviewed recently by Loarte eta/.10