Divertor

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A divertor is a device in fusion reactors that removes the fusion product helium -4 and impurities from the fusion plasma . Divertors can be used in ring-shaped fusion reactors of the tokamak and stellarator types .

Divertor cassette in a model of the ITER tokamak , at the lower end of the D-shaped plasma vessel cross-section

functionality

In order to maintain a continuous fusion , the fusion product helium-4 as well as impurities that are inevitably released from the wall material (see sputtering ) must be removed from the fusion plasma during operation. Since the plasma consists of completely ionized atomic nuclei ( they no longer have electron shells ), each of the nuclei carries as many positive electrical elementary charges as corresponds to its atomic number . Thus, the 4-helium nuclei formed during the fusion have twice the charge of hydrogen , deuterium and tritium nuclei , and an iron nucleus , for example, has 26 times the charge. By means of suitable additional magnetic fields , all nuclei with more than one elementary charge are directed out of the plasma onto cooled baffle plates of the divertor. There they lose their energy and can thus trap electrons, thus becoming neutral atoms; these are removed from the plasma vessel by vacuum pumps . The light nuclei remain in the plasma despite these additional magnets.

The impact plates are exposed to a high thermal load due to the impact of the particles (in addition to the load on all parts close to the plasma from the fast neutrons ). Metals with a high melting point such as molybdenum and tungsten , but also graphite , can be used as the material . Composite materials made from these elements are also being tested. The exact shape of the plates is important so that only the impurities are separated from the plasma flow.

The divertor must also have a modular structure so that its parts can easily be exchanged if necessary. Because of the radiation from the activated parts, repairs and maintenance work would have to be carried out remotely after commissioning.

Development history

The limiter

In earlier test systems (e.g. JET ), attempts were made to limit the plasma only with the so-called limiter and to capture all particles that are located behind the last closed magnetic surface. The limiter consists of plates that protrude more or less far into the reactor space, depending on the desired plasma limit. It is designed in such a way that the heat released by it can be absorbed without any problems. However, it was found that the atoms (e.g. iron, nickel, chromium, oxygen) released from the limiter itself due to the high load led to strong energy losses in the plasma, which were noticeable in the form of emitted light.

The magnetic limiter

Experiments with a limiter based on magnetic fields that avoid contact of the plasma with the surrounding wall were more promising. The attempt is made to separate the area of ​​the "good" plasma (with only minor contamination) from the outer area by means of a special shape of the magnetic field. The separation takes place on the enveloping surface in which the outermost closed magnetic field lines run, the “separatrix”. Since most of the impurities arise on the wall itself, they are directed back onto the wall by the open field lines outside the separatrix and do not penetrate into the plasma. Because of the special shape of the magnetic fields, which limits the plasma, this is called a "magnetic limiter". Since direct contact between the plasma and the wall is avoided in this process, the edge of the plasma and thus also the plasma can be much hotter. The associated improvement in magnetic confinement has been demonstrated in many experiments.

The divertor

However, the best results in terms of containment time and cleanliness of the plasma are obtained with divertors . The resulting impurities are not simply transported back onto the wall, but rather directed onto specially equipped panels. The neutral gas produced during the neutralization at the so-called divertor plates , which has a higher pressure than the main plasma flow, is conveyed out of the reactor chamber by divertor pumps. Due to the convincing results regarding the energy containment time and cleanliness of the plasma with ASDEX (axially symmetrical divertor experiment) and ASDEX upgrade , future fusion reactors ( ITER , DEMO , Wendelstein 7-X ) will be equipped with divertors. The basic functional application of the divertors will depend on the design of the fusion reactors.

The divertor in the tokamak

In 1981 the first divertor arrangement (Divertor I) in a tokamak was successfully tested with the help of experiments at ASDEX . A plasma state with very good thermal insulation, the so-called H regime (high confinement regime), was found. This discharge process is still the basic option for future fusion reactors today. Gradually, divertors were used in all tokamak-type reactors (ITER, JET, ASDEX, ASDEX Upgrade, etc.). In all of these experimental reactors, the divertor plates are arranged symmetrically on the floor in the entire reactor space. With the help of special magnetic fields, the plasma is separated from impurities and the "fusion ash" (helium-4) with the impurities is directed onto the divertor plates. Under conditions similar to that of a power station, the very strongly bundled contaminated plasma particles cause an extremely high thermal load on the plates. Attempts are made to handle this heat output through various concepts. One of these concepts is based on the targeted contamination in the edge area of ​​the fusion chamber by blowing in the noble gas neon. In the edge area, the plasma, which is still very hot, loses part of its energy and passes it on to the neon atoms, which consequently emit light in the ultraviolet or X-ray range. In contrast to the interior of the plasma, where this cooling would have to be avoided, here it reduces the power hitting the divertor plates in the edge area.

Another concept is based on an improved geometric arrangement of the baffle plates. An attempt is made to guide the neutral gas produced when the ions hit it in such a way that it absorbs part of the energy of the subsequent plasma. Experiments showed that the impact plates were noticeably relieved. Impurities could be directed in such a way that they concentrated in the divertor area and did not drift into the interior. This so-called "Divertor II" was first installed in ASDEX Upgrade in the summer of 1996 and modified as "Divertor II b" in the autumn of 2000 in such a way that an increased "triangularity" of the plasma was achieved. This procedure can further relieve the divertor plates. In addition, various divertor materials were tested to withstand the power, particle and heat flows under future power plant conditions. A tungsten coating on the divertor plates as well as on the inner wall of the reactor proved to be superior to the graphite previously used because of its thermal and mechanical properties .

The divertor in the stellarator

In 1994, the first preliminary studies for a stellarator divertor were started at Wendelstein 7-AS . The graphite limiter bricks used in earlier stellarators as well as in the tokamak had inevitably overheated during high heating outputs and long discharges (continuous operation of the stellarator). Here, too, magnetic fields were used to guide the helium -4 and the impurities generated during the fusion process onto baffle plates and to separate the inner part of the plasma from the outer part using the separatrix. In the case of the stellarator, no additional divertor magnetic field has to be generated, as “magnetic islands” exist in its non-axially symmetrical field. For better control of the plasma-wall interaction through the separatrix, the divertor modules were installed at the positions of these magnetic islands. As with the tokamak, the contaminated plasma to be discharged hits the baffle plates and can be discharged there by the divertor pumps. Several experiments have already shown largely stationary plasma discharges over long energy and particle confinement times with a very high plasma density and maximum heating power.

Current research

Since the divertor plates are among the most thermally stressed components in a fusion reactor (approx. 17% of the fusion power, which corresponds to peak loads of approx. 15 to 20 MW / m²), the main focus of the current (2008) research is on the development of appropriate materials and the geometric arrangement of the divertor modules. The divertor materials must have a very good heat transfer with high resistance to the plasma. Modules made of tungsten are currently being tested in the ASDEX upgrade under various conditions. With regard to the geometry, attempts are made to utilize the dynamic properties of the impinging plasma and the resulting neutral gas in such a way that the power / area impinging on the divertor plates is reduced. Furthermore, the divertor modules must be designed so that they can be replaced by robots in order to keep the radiation exposure of the personnel, especially in future power reactors , low.

See also

literature

Web links

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  1. ↑ Fusion product ( Memento of the original dated February 27, 2008 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.  @1@ 2Template: Webachiv / IABot / www.ipp.mpg.de
  2. ↑ Overview of ASDEX Upgrade results (O. Gruber et al. 2001 Nucl. Fusion 41 1369-1389)
  3. ↑ a b High heat flux performance of plasma facing materials and components under service conditions in future fusion reactors ( Fusion Science and Technology , 53 (2T): 278-287 FEB 2008)
  4. ↑ Modeling of radial electric field profile for different divertor configurations (V Rozhansky et al 2006 Plasma Phys. Control. Fusion 48 1425-1435)
  5. ↑ Chapter 8: Edge and divertor physics in ASDEX Upgrade ( Fusion Science and Technology , 44 (3): 659-681 NOV 2003)
  6. ^ R. Neu, M. Balden, V. Bobkov u. A .: Plasma wall interaction and its implication in an all tungsten divertor tokamak. Plasma Phys. Control. Fusion 49, B59-B70 (2007)
  7. ↑ Gadelmeier, F., P. Grigull, K. McCormick, ..., D. Hildebrandt, ..: Island Divertor Experiments on the W7-AS Stellarator. 29th EPS Conf. on Controlled Fusion and Plasma Physics, Montreux 2002. 29th EPS Conference on Controlled Fusion and Plasma Physics, Montreux, 2002. ECA VOL.26B (2002) (CD)
  8. ↑ Grigull, P., K. McCormick, J. Baldzuhn, R. Burhenn, R. Brakel, H. Ehmler, Y. Feng, F. Gadelmeier, L Giannone, D. Hartmann, D. Hildebrandt, M. Hirsch, R Jaenicke, J. Kisslinger, J. Knauer, R. König, G. Kühner, H. Laqua, D.Naujoks, H. Niedermeyer, N. Ramasubramanian, N. Rust, F. Sardei, F. Wagner, A. Weller , U. Wenzel and the W7-AS Team: First island divertor experiments on the W7-AS stellarator, Plasma Phys. Control. Fusion 43 No 12A (December 2001) 175-193
  9. ↑ Grigull, P., K. McCormick, Y. Feng, A. Werner, R. Brakel, H. Ehmler, F. Gadelmeier, D. Hartmann, D. Hildebrandt, R. Jaenicke, J. Kisslinger, T. Klinger, R. König, D. Naujoks, H. Niedermeyer, N. Ramasubramanian, F. Sardei, F. Wagner, U. Wenzel, and the W7-AS Team: Influence of magnetic field configurations on divertor plasma parameters in the W7-AS stellarator . 15th Int. Conf. on Plasma Surface Interaction in Controlled Fusion Devices, Gifu (Japan), May 2002
  10. ^ Degradation and Defects in Plasma Facing Components for Future Fusion Devices