Nuclear fusion reactor
A nuclear fusion reactor or fusion reactor is a technical system in which the nuclear fusion of deuterium and tritium takes place in a controlled manner as a thermonuclear reaction . Fusion reactors that would be suitable for generating electricity in a fusion power plant do not yet exist . Although this goal has been pursued since the 1960s, it is only slowly approaching due to high technical hurdles and also due to unexpected physical phenomena.
Research is currently (2020) focused on tokamaks and stellarators . These reactor concepts are based on the technique of magnetic confinement . A few grams of the deuterium-tritium gas mixture is placed in an evacuated, many cubic meter, toroidal container and heated to 100 to 150 million degrees Celsius. At these temperatures, electrons and atomic nuclei are separated from one another and form an electrically conductive plasma . Superconducting electromagnets that generate a magnetic field of up to 10 Tesla are arranged around the toroidal plasma chamber . This magnetic field traps the plasma in the chamber so that it does not touch the walls. If it came into contact with the wall, the plasma would cool down immediately and the reaction would collapse. The particle density corresponds to a technical vacuum . The strongly exothermic nuclear reaction occurs when the fast atomic nuclei collide. In the process, high-energy neutrons are released. The neutrons give off their energy in the blanket (outer jacket) as heat that is to be used to generate electricity.
The most important European research reactors are the tokamaks JET in Culham in Great Britain and ASDEX Upgrade in Garching near Munich as well as the stellarator Wendelstein 7-X in Greifswald . The most promising project is the international research reactor ITER , a tokamak that has been under construction in Cadarache in southern France since 2007 . The aim of ITER is to show that it is possible to generate technically usable energy through nuclear fusion. The generation of the first hydrogen plasma is planned for 2025. Operation with a deuterium-tritium plasma is not expected to take place before 2035 at the earliest. The knowledge gained with ITER should provide the basis for the construction of the demonstration power plant DEMO , which should breed sufficient tritium and reliably feed several 100 MW into the power grid.
Potential energy source of the future
With the development of nuclear fusion reactors it is hoped that a practically inexhaustible source of energy will be developed without the risk of catastrophic accidents and without the need for permanent storage of long-lived radioactive waste. If nuclear fusion reactors reach the technical maturity for power generation, a first commercial reactor cannot be expected before 2050 based on current knowledge. Large-scale use is foreseeable in the last quarter of the 21st century, provided the technology meets with acceptance and is economical. Today (2020) no statement can be made about the profitability. The future regional advantages and disadvantages compared to other electricity generation methods, the reactor construction or import costs, the expenses for financing, operation, dismantling and disposal of radioactive waste as well as the then valid electricity price cannot be reliably forecast.
During the development of the atomic bomb , Edward Teller , Enrico Fermi and other scientists presented the first drafts for generating electricity through controlled nuclear fusion. One concept envisaged the use of a magnetic field to enclose the deuterium-tritium plasma, which had to be heated to several million degrees for the fusion. After the Second World War, the first civil research program on the use of nuclear fusion was started on this basis in England. George Paget Thomson and Moses Blackman pursued the idea of the annular confinement of the plasma. High-frequency electromagnetic waves were provided for heating.
First stellarators and tokamaks
In the following years, this concept was further developed independently in two variants in the USA and the Soviet Union. In the USA, Lyman Spitzer developed the stellarator , the behavior of which was researched from 1951 as part of the Matterhorn and Sherwood projects at Princeton University, among others .
A magnetic field should serve to contain the particles, in which field lines for the magnetic containment run within nested torus surfaces. It soon became apparent that such flux areas are not easily accessible in the stellarator. The theoretical basis for this was only developed gradually. Only towards the end of the 20th century could the necessary calculations be carried out thanks to sufficiently powerful computers; this made it possible to build the Wendelstein 7-X stellarator , which generated its first plasma in 2015.
In 1950 and 1951, Andrei Sakharov and Igor Tamm tried another variant of magnetic confinement in the Soviet Union , the tokamak . According to this concept, a magnetic field generated in the plasma itself by the flow of current is involved in confinement; the current in the plasma also helps to heat it. A surprising temperature record was set in the Soviet T3 tokamak in 1968 with 10 million ° C over 10 milliseconds. After this became known in the West, the simpler tokamak design became the basis of almost all subsequent relevant experiments.
First successes in the EU and the USA
The first attempts to generate energy from nuclear fusion had taken place independently of one another and under military secrecy. In 1956, Igor Wassiljewitsch Kurchatov , the former head of the Soviet atomic bomb program, broke the secrecy with a lecture at the Harwell research center in England . At the second international Geneva nuclear conference in 1958, it was decided for the first time to disclose the results and to strengthen international cooperation, also in view of the great technological difficulties.
In Europe, the Euratom Treaty was signed in 1958 , in which six countries initially committed themselves to work together in the field of nuclear energy and nuclear research. In 1973 it was decided to build the Joint European Torus (JET) in Culham (Great Britain), which is currently the largest tokamak. In 1983 the reactor went into operation. On November 9, 1991, JET was able to release a significant amount of energy from controlled nuclear fusion for the first time. A deuterium-tritium plasma provided an output of 1.8 megawatts for two seconds. In 1997, a fusion power of 16 megawatts was achieved, although 24 megawatts were required for the plasma heating.
Since the Soviet temperature record in 1968, American Princeton University had worked intensively on tokamak projects in addition to the stellarator concept. The Tokamak Fusion Test Reactor (TFTR) in the Princeton Plasma Physics Laboratory (PPPL) achieved similar successes as the competing European JET; In 1994, 10.7 megawatts of fusion power were achieved, and in 1995 a plasma temperature of 510 million ° C. The TFTR was in operation from 1983 to 1997. From 1999 to 2016, research was carried out on the successor National Spherical Torus Experiment (NSTX).
International projects and plans
A large number of technical difficulties still have to be overcome in a wide variety of fields before the first practical, continuously operating and economically profitable fusion reactor can be found. The development of the civil use of fusion energy is also being promoted in international projects because of the high costs. The magnetic containment method is used almost exclusively worldwide .
In a nuclear fusion, atomic nuclei fuse to form a new nucleus. Many nuclear reactions of this type release energy. The energy emitted by the sun comes from nuclear fusion processes. In its center, hydrogen fuses in the proton-proton reaction and in the CNO cycle under a pressure of 200 billion bar at around 15 million degrees Celsius to form helium . However, because of the extreme pressure, these processes are unsuitable for use on earth.
In order for a fusion reaction to occur between two atomic nuclei, they have to come very close to one another, to around 2.5 femtometers (see Strong nuclear force ). This is counteracted by the electrical repulsion , which must be overcome with a large expenditure of energy (high temperature). The fusion reactions suitable for technical energy generation are well known from investigations using particle accelerators . In accelerator experiments, however, much more energy is expended to operate the apparatus than the reaction then releases; a net energy gain, i.e. the operation of a power plant , is not possible in this way.
In order for a nuclear fusion to be able to convert matter into energy according to Einstein's formula E = mc 2 , the mass of the two fusing nuclei together must be greater than the mass of the nuclei and particles formed. This mass difference is converted into energy. The mass difference is particularly large when helium -4 is formed from isotopes of hydrogen . These also have the smallest electrical repulsion to be overcome before the fusion, because they each carry only a single elementary charge. A mixture of equal proportions of deuterium (D) and tritium (T) is therefore intended as a fusion fuel :
This reaction is also characterized by an effective cross-section which characterizes the reaction probability and which is sufficiently large even at plasma temperatures that can just be technically achieved. All realistic concepts for fusion power plants are based on this reaction until today (2016).
Fusion with magnetic plasma confinement
The most promising concepts for fusion reactors to date envisage enclosing a deuterium-tritium plasma in a ring-shaped magnetic field and heating it to a sufficient temperature. In order to achieve a net energy gain in this way, the plasma volume must be sufficiently large (see A / V ratio ).
To get the process going, a few grams of a deuterium-tritium gas mixture (1: 1) are let into the many cubic meters large, well evacuated reaction vessel; the particle density then corresponds to a fine to high vacuum . The gas is brought into the plasma state by heating and is further heated. After reaching the target temperature - around 150 million degrees in the innermost part of the plasma - the plasma exerts a pressure of a few bar . Against this pressure, the magnetic field has to hold the particles together. Contact with the vessel wall must be prevented, otherwise the plasma would cool down immediately.
Fusion reactions take place at a temperature of approx. 150 · 10 6 ° C and a particle density of approx. 10 20 m −3 . The energy released is distributed as kinetic energy in a ratio of 1: 4 to the alpha particles (He-4 nuclei) and free neutrons (see kinematics (particle processes) ). The energy of the alpha particles is distributed further through collisions in the plasma and contributes to its further heating. With a sufficient nuclear reaction rate (number of reactions per time interval), this energy can be sufficient to maintain the plasma temperature without further external heating: The plasma then "ignites" and "burns" by itself. This occurs when the triple product of the particle density at a given temperature , Temperature and a time constant determined by the inevitable heat losses, the energy containment time, exceeds a certain minimum value according to the Lawson criterion .
However, this point does not have to be reached for an energy-supplying reactor. Even at slightly lower temperatures and constant additional heating, enough fusion reactions take place (see fusion with net energy gain without reaching the Lawson criterion ). The additional heating even offers a welcome option (in addition to refilling the fuel) to control the reaction rate, i.e. the reactor output. The plasma state that has been reached must be permanently maintained by topping up new fuel according to consumption and removing the helium that is created - the result of the fusion, the "ash". The released neutrons leave the plasma; their kinetic energy, four fifths of the fusion energy, is available for use.
An energy gain has so far only been achieved very briefly in tests at JET and TFTR (Princeton, USA), but not yet in the many other experiments, because the plasma vessels of the existing test facilities are too small, which means that the plasma cools down too much (see A / V ratio ). In the ITER tokamak, which is therefore larger, a permanent “burning” fusion is to be implemented with constant additional heating. Even later systems like DEMO will probably be designed in such a way that a weak additional heating of, for example, a few percent of the fusion power remains necessary in order to retain an additional control option.
Increasing the temperature or density increases the power produced by fusion reactions . It is not possible, however, to increase the temperature to too high, since the energy loss of the plasma due to transport processes also increases with the temperature. The desired reaction rate thus remains constant at constant temperature and density.
Various methods have been developed for heating the plasma to over 100 million ° C. All particles in the plasma move at a very high speed according to the respective temperature (deuterium nuclei at 100 million ° C have an average speed of around 1000 km / s). The heating power increases the temperature and compensates for the losses caused by mainly turbulent and neoclassical transport (caused by collisions between the particles) and by bremsstrahlung .
With some of the following heating methods, the temperature and thus also the current distribution in the plasma can be influenced, which is important for its dimensional stability:
- Electrical heating: The plasma is an electrical conductor and can be heated by means of an induced electrical current. The plasma is the secondary coil of a transformer . However, the conductivity of the plasma increases as the temperature rises, so that the electrical resistance from around 20-30 million ° C or 2 keV is no longer sufficient to heat the plasma more strongly. With the tokamak, the current through the central solenoid is continuously increased for electrical heating.
- Neutral particle injection : When fast neutral atoms are injected into the plasma ( neutral beam injection , NBI for short), the kinetic energy of these atoms - which are immediately ionized in the plasma - is transferred to the plasma through impacts, which causes it to heat up.
- Electromagnetic waves: Microwaves can excite the ions and electrons in the plasma at their resonance frequencies (rotational frequency in the helical line that describes the particle in the magnetic field) and thus transfer energy into the plasma. These methods of heating are called Ion Cyclotron Resonance Heating (ICRH), Electron Cyclotron Resonance Heating (ECRH) and Lower Hybrid Resonance Heating (LHRH).
- Magnetic compression: The plasma can be heated like a gas by rapid ( adiabatic ) compression. An additional advantage of this method is that the plasma density is increased at the same time. Only magnetic fields generated by magnetic coils with variable current strength are suitable for compressing the plasma.
The magnetic field must hold the plasma together against its pressure so that it does not touch the vessel wall. Both concepts for magnetic confinement, tokamak and stellarator , use a toroidal , twisted magnetic field. Tokamaks generate the twisting of the field by inducing an electric current in the plasma, stellarators accomplish this by a special, complicated shape of their magnetic coils (more detailed explanation of the magnetic confinement and the necessity of twisting the field lines in fusion by means of magnetic confinement ).
Special, localized deformations of the field remove the unwanted ions, i.e. the fusion product helium and any impurities, from the plasma (see divertor ).
The magnetic field is generated with large coils. Their shape and arrangement determine the shape of the plasma; The strength of the current in the coils determines the strength of the magnetic field and thus the possible size of the plasma, the particle density and the pressure. In a reactor (or in experiments in which the plasma is confined for a longer period of time) the coils must be superconducting: those in normally conducting coils flowing electricity produces heat due to the electrical resistance to be overcome. Such coils could no longer be effectively cooled if they were operated for a longer period of time, as a result of which the temperature would rise and the coil would be destroyed. Superconducting coils, on the other hand, have no resistance, which is why the current in them does not produce any heat that has to be dissipated.
The tokamak is the most advanced and internationally pursued concept with ITER. However, at least in its original mode of operation with a purely inductively generated plasma flow, it has the disadvantage that operation is not possible continuously, but only in pulsed mode, that is, with regular short interruptions. Therefore be
- on the one hand other, additional possibilities for "driving" the current in tokamaks developed,
- on the other hand, stellarators continue to be pursued as an alternative.
Experiments on the ASDEX upgrade and other research reactors indicate that tokamak reactors could work continuously in the future.
Occurrence and procurement
While deuterium is present in almost inexhaustible quantities (2.5 · 10 13 t) in the water of the earth , tritium in the quantities required for a fusion reactor can practically only be produced by "incubating" lithium -6 in the plant itself:
The earthly occurrence of lithium is estimated at more than 29 million tons. Only the isotope 6 Li, which occurs naturally at 7.5%, is used for tritium incubation . From this proportionate supply of around 2 million t of lithium-6, around 1 million t of tritium can theoretically be obtained using the above formula. In practice, enriched lithium with a lithium-6 content of 30 - 60% should be used. The technically usable lithium deposits are mathematically sufficient to meet the energy needs of mankind for thousands of years.
A shortage due to the need for lithium in other branches of industry stands in the way of the fact that the isotopic composition does not play a role in these branches and therefore over 90% of the lithium remains available for them. Even in a scenario with a sharp increase in lithium demand due to the massive expansion of electromobility, by 2050 only those lithium resources that can be mined at today's lithium prices and technologies will be exhausted.
Tritium is radioactive with a half-life of 12.32 years. However, it only emits beta radiation with a low maximum energy and without accompanying gamma radiation . In the radioactivity inventory of a fusion reactor that has been in operation for some time, tritium will only make a relatively small contribution.
The tritium required to start fusion reactors could easily be obtained in conventional nuclear fission reactors. In particular, heavy water reactors (for example CANDU ) produce tritium as a by-product in an amount of around 1 kg per 5 GWa of electrical energy generated. The tritium (a few kilograms) required during the planned term of ITER is also likely to come from this.
For the continuous operation of fusion power plants, however, these hitherto only available sources would by far not supply enough tritium, which results in the need to breed tritium in the reactor. A fusion power plant with 1 GW of electrical output would require around 225 kg of tritium per year.
Tritium breeding and neutron multiplication
An economical production of the necessary amounts of tritium would be possible by the above-described production from lithium-6 in the fusion reactor itself by means of the free neutrons emitted anyway. For this purpose, the plasma is surrounded by a brood mantle, the blanket.
Nuclear fusion delivers exactly one neutron for each tritium atom used; In principle, a new tritium atom could be produced from this. However, this is not possible without loss, because purely geometrically, the blanket cannot capture 100% of the neutrons, and some of the neutrons hitting the blanket are inevitably absorbed in atomic nuclei other than lithium or escape from the system. Losses are inevitable when the tritium is transferred into the fusion plasma, as is its radioactive decay. In order to still be able to bring as much new tritium into the plasma as was used, the neutrons in the blanket must be increased by around 30 to 50%. For this purpose, the blanket drafts provide for the use of the (n, 2n) nuclear reaction either on beryllium or on lead . Commercial fusion reactors must therefore be designed in such a way that a slight overproduction of tritium is possible. The tritium breeding ratio can then be set and readjusted via the degree of enrichment of the 6 Li isotope in the blanket.
The technological development of this tritium extraction is a crucial task for future fusion research, especially at ITER. Whether the incubation of tritium is possible with sufficient efficiency in practice will only be seen when a first deuterium-tritium fusion reactor works with it in continuous operation. But only if the plants can cover their own tritium needs themselves and the quantities required to start a fusion process can be obtained elsewhere, it is possible to build up a power supply using fusion reactors. This question is discussed in scientific publications. While some scientists such as Michael Dittmar from CERN criticize the self-sufficiency of fusion reactors with tritium as unrealistic in view of previous experimental and computational results, most fusion researchers do not see any fundamental problems on this point.
During the burning time of the plasma, it is necessary to refill fuel according to the respective consumption. For this purpose, the shooting of pellets from a frozen deuterium-tritium mixture into the vessel has proven to be a suitable technique. Such pellets with a mass of, for example, 1 mg are brought to a speed of about 1000 m / s by a centrifuge or pneumatically with a type of gas gun. This refill method also makes it possible to specifically influence the spatial density distribution of the plasma through the choice of the injection point and the pellet speed. With more or less refills, the rate of fusion can also be controlled; stopping the refill ends the fusion reaction.
Removal of helium and contaminants
The reaction product 4 He and cores inevitably knocked out of the wall material act as impurities; they must be constantly removed from the plasma. Since they have higher numbers of charges than the hydrogen isotopes, this is possible with magnetic deflection. Specially developed divertors are used for this ; They consist of baffle plates mounted on the edge of the torus, onto which the ions undesired in the plasma are directed with an auxiliary magnetic field. There they cool down and thereby capture electrons again, i. that is, they become neutral atoms. These are not influenced by the magnetic field and can be extracted by the suction system that maintains the high vacuum.
Use of the released energy
From the energy yield of the nuclear reaction, 17.6 MeV per individual reaction, four fifths, i.e. 14.1 MeV, occur as kinetic energy of the released neutron. These neutrons are hardly influenced by the magnetic field and get into the blanket, where they first give off their energy as usable heat through impacts and then serve to incubate one tritium atom each. The thermal energy can then, as in any conventional power plant through heat exchanger generate steam, which in turn steam turbine with coupled power generators drives.
The useful energy of the deuterium-tritium reactor occurs in the form of high-energy neutrons (14.1 MeV ). The neutrons hit the plasma-facing side of the blanket with a high flux density , around 10 14 s −1 cm −2 - in addition to being exposed to thermal radiation. This inevitably leads to considerable radiation damage in the material (compared to self-right in the core of a typical pressurized water reactor , the neutron flux density is around ten times smaller, and are there predominantly to thermal neutrons). The radiation damage depends heavily on the energy of the neutron. That is why the wall load is often given as the product of neutron flux density and neutron energy, i.e. as the power surface density in MW / m² ( megawatts per square meter). With an energy of 14.1 MeV, 10 14 neutrons s −1 cm −2 correspond to about 2.2 MW / m². This is the neutron wall loading provided for in a draft for the blanket of the DEMO reactor. The blanket should have a service life of 20,000 operating hours, i.e. around 2.3 years. The accumulated dislocation damage - which mainly causes embrittlement - amounts to about 50 dpa (displacements per atom) in steel . In addition, the material is damaged by swelling because (n, p) and (n, α) nuclear reactions in the metal structure produce gases, hydrogen and helium, respectively. Helium in metal is also detrimental to weldability . A helium concentration below 1 appm ( “atom parts per million” , i.e. one He atom per 1 million metal atoms) is required so that steel parts and pipe connections can be welded together again after being replaced .
In addition, radioactive nuclides are formed in the materials through activation . In order to produce the smallest possible quantities of it, which should also have the shortest possible half-lives , only materials made from certain elements can be used. In today's common structural materials such as austenitic chromium-nickel stainless steels , neutron activation produces large quantities of the relatively long-lived and strongly gamma-emitting 60 Co. The structural material of ITER is still an austenitic chromium-nickel steel; Such steels, however, cannot be used for future power plant reactors.
The main requirements for material development are materials that can be activated with low levels of activity, that have sufficient resistance under neutron radiation and that must meet all the requirements for their respective special tasks, such as stability, amagnetism or vacuum tightness. So far it has also been assumed that the innermost shell must be replaced periodically, since no material will withstand the high neutron flux of a commercial reactor for years. Because of the radiation from the activated parts, repairs and maintenance work would have to be carried out remotely after commissioning. The aim is to ensure that the majority of the capitalized plant components have to be stored in a controlled manner for only about 100 years after the end of their useful life, until recycling is possible; the smaller part has to be stored for about 500 years. A disposal would therefore not necessary. The development work focuses on nickel-free, ferritic-martensitic steels, but vanadium- based alloys and ceramic silicon carbide (SiC) are also investigated. With ASDEX Upgrade it was found that tungsten is also suitable for the front surfaces of the blanket modules facing the plasma and for divertor plates . For irradiation experiments on these materials, the high-intensity and high-energy neutron source IFMIF is to be operated at about the same time as ITER .
A spatially detailed calculation of the activation in a DEMO reactor was presented in 2002 by the Karlsruhe Research Center . A fusion power of 2200 MW was assumed for the reactor. Its blanket consists of 77 t (tons) of lithium orthosilicate Li 4 SiO 4 (enriched to 40% lithium-6) as a breeding material, 306 t of metallic beryllium as a neutron multiplier and 1150 t of the Eurofer steel currently in development (main components 89% iron, 9% Chromium and 1.1% tungsten) as structural material. For all materials, not only the nominal, ideal composition was taken into account, but also the typical natural impurities, including, for example, a proportion of 0.01% uranium in beryllium. The activity was calculated at the end of an uninterrupted full load operation of 20,000 hours; this is the service life required for the DEMO blank parts until replacement. The gamma radiation dose rate on the material surface of a solid component was considered to be the determining variable for later handling of the activated parts . It was assumed that reprocessing into new reactor parts is possible at less than 10 mSv / h (millisievert per hour) with remote control technology (remote handling) and at less than 10 Sv / h with direct handling (hands-on handling) . The result is that all materials - lithium silicate, beryllium and steel - can be remotely processed after a decay time of 50 to 100 years. Depending on its exact composition, it can take up to 500 years for steel to fade to direct manageability.
In 2006, the total amount of radioactive material accumulating during a 30-year lifetime of a facility was estimated to be between 65,000 and 95,000 tons, depending on the type of construction. Despite this greater mass , their activity in Becquerel would be comparable to the dismantling products of a corresponding fission reactor; however, the environmental properties would be significantly more favorable. In contrast to nuclear fission power plants, neither large amounts of fission products were left during electricity production, nor ore residues that produce radioactive radon.
state of research
In nearly 50 years of fusion research [out of date] since the results with the first Russian tokamak T3 from 1968, each of the three decisive variables - temperature , particle density and energy inclusion time - has been increased considerably and the triple product has already been improved by a factor of around 10,000; it is still about a factor of seven away from ignition, for which the triple product must have a value of about 10 21 keV s / m³. In smaller tokamak systems, the temperatures reached have already been increased from 3 million ° C to over 100 million ° C.
The main goal of current research on the two magnetic confinement methods is to find plasma conditions that significantly increase the energy confinement time . In the many previous experiments, the measured energy containment time has proven to be much shorter than theoretically expected. At the end of April 2016, the Max Planck Institute for Plasma Physics reported that experiments at the ASDEX Upgrade with regard to the inclusion time were successful and that continuous operation of a tokamak is technically feasible. This means that the “conditions for ITER and DEMO are almost fulfilled”.
The Wendelstein 7-X stellarator, which was completed in 2015 , initially only works with hydrogen; later, deuterium will also be used. It is intended to demonstrate the continuous permanent plasma confinement without current flow in the plasma - the main advantage over tokamaks. This would show that the stellarator concept is also fundamentally suitable as a fusion power plant.
The previous systems are still too small to ignite the plasma, so that the plasma cools down too much. A certain minimum size of the plasma is necessary in order to reach 10 to 15 keV (110–170 million ° C) in the center, because with a given size the plasma can only have a certain maximum total energy. A positive energy balance is to be achieved for the first time in the future international fusion reactor ITER, which has been built since 2007 in the Cadarache research center in the south of France . The reactor should deliver around ten times more fusion power than has to be used to heat the plasma. For this purpose, the temperatures required for such fusion rates are to be generated from 2026 with additional heating . At ITER, tritium breeding and the necessary neutron multiplication are also to be developed and optimized. The research results from ITER are intended to pave the way for the first "demonstration power plant " DEMO , which will generate electricity from 2050 and thus demonstrate the commercial viability of nuclear fusion.
No other merger concept has reached a level of development that, from today's perspective (2019), can be considered for electricity generation.
- Fuels other than deuterium-tritium would pose far greater technical difficulties. Only in experimental facilities for plasma physics, in which energy production is not the goal, is it worked with pure deuterium in order to avoid the practical complications caused by the radioactive tritium.
- The concept of inertial confinement is at the basic research stage, is not primarily geared towards the development of power plants and is far from commercial use.
- In the opinion of most scientists, cold fusion is also not a possible alternative. Technical processes of this kind with the claimed release of energy are not possible according to known physics.
List of test facilities
The most important systems are listed in the following table.
|Finished experiments||Systems in operation||Construction in progress|
|Tokamaks||Tokamak Fusion Test Reactor (TFTR) at Princeton University , USA (1983–1997)||Joint European Torus (JET) in Culham , England||ITER in Cadarache , France|
|National Spherical Torus Experiment (NSTX) at Princeton University, USA (1999-2016)||ASDEX upgrade at the Max Planck Institute for Plasma Physics in Garching near Munich|
|TEXTOR at the Institute for Plasma Physics at the Research Center Jülich (1983–2013)||Experimental Advanced Superconducting Tokamak (EAST) in Hefei, China|
|JT-60 in Naka, Japan|
|Tokamak à configuration variable (TCV) from the Swiss Federal Institute of Technology in Lausanne , Switzerland|
|Tore Supra / WEST in Cadarache, France|
|KSTAR in Daejeon, South Korea|
|Stellarators||Wendelstein 7-AS in Garching near Munich (1988–2002)||Wendelstein 7-X in Greifswald|
|National Compact Stellarator Experiment (NCSX) at Princeton University, USA (2003-2008, construction not completed)||Columbia Non-Neutral Torus at Columbia University in New York , USA|
|Large Helical Device (LHD) in Toki (Gifu), Japan|
|H-1NF in Canberra, Australia|
|TJ-II at CIEMAT in Madrid, Spain|
|National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory in Livermore, California , USA|
|National Laser Users' Facility (NLUF)|
|Laser Mégajoule in Le Barp, South West France|
|Dense Plasma Focus|
|ECRIS driven neutronless fusion|
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