High-temperature reactor concepts use helium gas as a coolant instead of water. This allows the reactor to operate at lower pressure, making it more controllable at extremely high temperatures compared to conventional light water reactors. Uranium oxide or carbide is predominantly used as fuel. The fuel comes in small pellets that are encased in a protective shell. The pellets, in turn, are embedded in spheres or prismatic blocks of graphite, which serves as a moderator. These spheres or blocks represent the fuel elements. Coolant flows around them and absorbs the heat generated during the nuclear reaction. This heat can be used, for example, to heat water and drive a steam turbine.

In addition to an increased efficiency and the generation of process heat at high temperatures, high-temperature reactors offer further advantages over conventional reactors. The design of the fuel elements and the helium cooling offer improved safety features. This means that additional safety systems can be used, some of which are not available in water-cooled reactors. Due to its design, the high-temperature reactor has a relatively low output in relation to the total volume of the reactor core. A core meltdown can, therefore, be ruled out.

If the plant is suitably designed, natural uranium, thorium, plutonium or mixed oxides can also be used as fuel in addition to enriched uranium.

However, the technology also has major disadvantages. The high temperature and the helium coolant pose a challenge in terms of selecting suitable materials. Gas-cooled reactors also often exhibit problems such as uneven cooling, high abrasion and dust formation as well as an increased risk of fire in the event of water or air ingress. This can lead to the release of radioactive substances.

Due to the high content of radioactive graphite, the final disposal of spent fuel elements is estimated to be significantly more cost-intensive compared to conventional fuel elements.

Gas-cooled high-temperature reactors have been the subject of research since the 1960s. Prototype plants based on this concept (the pebble bed reactors in Jülich and Hamm-Uentrop) were also developed in Germany. At the end of the 1980s, both plants were shut down due to various technical problems, and the technology was gradually abandoned in Germany. Other high-temperature reactor projects have been and continue to be developed in the UK, the USA, Japan and France, among others. A project in South Africa, which was based on AVR Jülich technology, was paused indefinitely due to technical difficulties and a lack of funding in 2010. A high-temperature experimental reactor, the HTR-10, which is also based on the pebble bed design, has been in operation in the People's Republic of China since 2003. Two further high-temperature reactors of the HTR-PM type there reached criticality as demonstration plants in autumn 2021. A similar project in the USA was discontinued before a prototype reactor was even built, but research on the high-temperature reactor concept is ongoing there. A general trend towards moderately high operating temperatures of 700-850°C can be observed in current developments.

To date, there is no high-temperature reactor for commercial power generation in operation.

The fuel is a mix of molten salts (fluorides and chlorides). The concentration of the fissile fuel can be adjusted very accurately via the selection of the salts and their mixing ratio. This allows the production of the exact concentration required to maintain a stable chain reaction. The temperatures in the molten salt are approx. 600-700°C. Controlled nuclear reactions that generate heat take place inside the reactor. This heat can be used to heat water vapour and power a turbine for electricity generation.

The safety concept of molten salt reactors is based on basic physico-chemical properties and requires less active safety technology than conventional light water reactors, for example. A central feature of the safety concept is to drain the molten salt into designated containers in the event of malfunctions, thus preventing any further chain reaction.

In addition, molten salt reactors can integrate what is known as chemical treatment. The fission products and the composition of the fission products, the fuel and the salt mixture used can be optimised during operation in an additional system in the primary circuit (fuel processing system). In contrast to light water reactors, there is no increased pressure in the primary circuit of a molten salt reactor, which means that some accident scenarios can be ruled out.

A major disadvantage of the molten salt reactor is the increased corrosion inside the pipe systems. The hot fuel-salt mix corrodes the metals in the reactor, thus limiting their service life. This problem is also the subject of current research and an important reason why, to date, molten salt reactors only exist as research or pilot plants.

Some concepts for molten salt reactors advertise the fact that they can also recycle radioactive waste. The idea is that so-called transuranium elements, which are produced in the reactor during nuclear fission, as well as individual long-lived fission products can be specifically converted, i.e. transmuted. This has not yet been developed to the point where it is ready for use. According to the current state of research, however, it would not be possible to convert all of the radioactive waste. New fission products would also be generated. There would, thus, be no advantage in terms of the final storage strategy pursued in Germany.

Depending on the specific design of the molten salt reactor concept, radioactive residues would be produced that differ from those of previous light water reactors. The entire disposal chain would have to be adapted, from the development of suitable conditioning processes and new containers to the requirements for interim and final storage of the radioactive residues.

Molten salt reactors were last operated in the USA in the 1950s and 1960s in the form of two experimental reactors. Research into the further development of this technology is currently underway in several countries. This research is at very different stages and includes concept studies as well as theoretical and experimental preliminary work. The development of an experimental reactor in China (TMSR-LF1) is the most advanced such concept. The commissioning of this reactor, which has been under construction since 2018, was approved by the Chinese authorities in summer 2022.

The supercritical-water-cooled reactor is a nuclear reactor that uses supercritical water as a working medium. The water is always in a supercritical state, i.e. it has a temperature of over 374°C and a pressure of at least 221 bar. No phase transitions take place above this point, known as the ‘critical point’ of water, which means that the water will no longer boil or condense.

The structure of the reactor corresponds to that of a boiling water reactor. The water in the reactor core is heated in a simple cooling circuit, and then fed directly into the turbine. Unlike in a boiling water reactor, the water does not vaporise in supercritical state. The coolant has a higher density and can, thus, absorb the heat more efficiently and transport it away from the core. The core temperature is higher than that of boiling and pressurised water reactors, and the pressure is significantly higher than that of pressurised water reactors (usually a maximum of 160 bar).

The design of the reactor is simple and the efficiency is high (up to 45 %). The special neutron spectrum of the supercritical light water reactor has fast neutrons as well as thermal neutrons. These cause long-lived radionuclides to be transmuted into shorter-lived ones, meaning that the spent nuclear fuel will radiate for less time.

One disadvantage is that, similar to the boiling water reactor, the turbine gets radioactively contaminated through direct contact with the cooling water in the primary circuit. The pressure in the circuit (approx. 250 bar) is very high, which is why the reactor pressure vessel and all other components of the primary circuit have to be thicker and more stable than in conventional light water reactors. Due to the high pressure, damage to the primary circuit also poses an increased risk.

The operation of coal-fired power plants with supercritical water was first trialled in the 1950s and is now standard in new construction projects. Research into the transfer of the concept to nuclear technology has been intensified since the 1990s. However, materials used in modern coal-fired power plants do not have sufficient corrosion resistance for use in the nuclear sector. Further relevant research and development into cladding and structural materials and safety functions is needed.

At present, the most advanced designs come from China, the EU, Japan, Canada, Korea, Russia and the US. On the whole, however, development is at an early stage. There are currently no plans for a prototype system.

The design of the reactor is similar to that of a classic pressurised water reactor (light water reactor). But instead of water, helium (other gases are also conceivable) is used as a coolant. Uranium, thorium, plutonium or compounds thereof are used as fuel. Unlike high-temperature reactors, which work with moderated thermal neutrons like conventional light water reactors, the fuel in fast reactors is split with fast neutrons. This means that the use of a moderator is not necessary. The high operating temperature of around 850°C yields high efficiencies or can be utilised as process heat for industrial processes.

The envisaged design of the reactor is relatively simple, and there is no need for a moderator at all. The use of unmoderated neutrons leads to transmutation, resulting in less long-lived nuclear waste. Moreover, helium as a coolant can be heated to very high temperatures and does not become radioactive itself.

This is the drawback of fast gas-cooled reactors, as helium is not very thermally conductive, which results in increased requirements for cooling the reactor core during operation and immediately after shutdown. Due to the high temperatures, only particularly heat-resistant materials can be used. An additional stress arises from the high neutron flux. The unmoderated fast neutrons are more difficult to shield and can penetrate further into materials than moderated neutrons. This impairs the service life of these materials.

Work on the fast gas-cooled reactor concept has been ongoing in the US and Germany since the 1960s, and later also in the UK and Japan. Since the 2000s, research has primarily been driven by France. So far, however, no helium-cooled fast reactor has been built and operated.

Extensive research and development are still required, particularly to find suitable fuels and cladding and structural materials for the high-temperature design. In addition, many questions regarding the necessary safety systems and the safety and reliability of operation in general remain unanswered. Generally speaking, development is still at the applied research stage, with no existing prototype designs. Commercial utilisation for power generation or industrial applications is not foreseeable.

The reactor core containing the fuel is located in a pool-type container filled with liquid sodium. Sodium is used for its high thermal capacity and good conductivity. Sodium does not boil during operation, so there is no elevated pressure in the reactor vessel. A heat exchanger inside the reactor vessel transfers the heat from the main circuit sodium to a secondary circuit, which also contains liquid sodium. From this secondary circuit, the heat is transferred to a water-bearing tertiary circuit that drives a turbine to generate electricity.

In contrast to many other reactor concepts, fast reactors use unmoderated fast neutrons. They can produce additional fissile material from non-fissile isotopes such as uranium-238 or thorium-232 during breeding reactions. Following reprocessing, the fissile material produced in this way can be used as nuclear fuel.

Another promise is the reduction of long-lived nuclear waste through transmutation, provided the reactor and fuel production are designed accordingly.

Thanks to its excellent heat capacity, sodium can completely absorb the decay heat of the fuel elements even without circulation. If, for example, the cooling system should fail due to a power failure, a core meltdown would be passively prevented. In the event of a leak, less coolant will escape as the primary and secondary circuits operate without pressure. This should result in advantages in terms of safety.

However, specific accident risks such as sodium leaks and fires must be considered. In the event of a coolant leak, it is necessary to prevent the highly reactive sodium from coming into contact with water and oxygen. This requires additional safety barriers. The system is complex and comparatively expensive, not least because it requires three cooling circuits.

Earlier decades saw the possibility of incubating additional fuel in reactors (breeder reaction) as an advantage in some cases. However, due to the quantity of uranium deposits worldwide, there were no major economic advantages to such an application. In addition, depending on the configuration, weapons-grade plutonium is incubated in the reactor. This increases the risk of proliferation of nuclear weapons-grade material.

With regard to the transmutation of long-lived waste materials, it must be noted that no such application has yet been developed to operational maturity. According to the current state of research, it would not be possible to transmute all of the radioactive waste. In addition, new fission products would be produced. This would therefore not be an advantage for the final storage strategy pursued in Germany, for example.

The sodium-cooled fast reactor was one of the first reactor concepts in the early days of civil nuclear energy utilisation. Sodium-cooled breeder reactors were and are in operation in several countries. One such experimental facility, the KNK-II, was operated at the German research centre in Karlsruhe from 1977 to 1991. The Kalkar nuclear power plant, which was based on the same technology, was never put into operation due to safety concerns.

Three fast sodium-cooled reactors are currently in commercial operation in Russia and China, and others are under construction in both countries and in India. Research and development of reactor concepts for this technology line is ongoing in a large number of countries around the world.

The "Generation IV International Forum" has given top priority to this development project. The plan is to press ahead with the development of an advanced fast sodium-cooled reactor with the option of transmuting particularly long-lived waste materials, and to move on to a trial phase in the 2020s. China, EURATOM, France, Japan, Korea, Russia and the USA are contributing to the research and development work.

The reactor has a pool design, which means that the reactor core is located in a pool-shaped container. The pool is filled with the coolant, which is either liquid lead or a lead-bismuth alloy. The metallic coolant does not boil during operation, meaning that normal pressure prevails in the reactor vessel. The heating and cooling processes in the various zones of the reactor vessel allow the coolant to circulate naturally without the need for pumps. A heat exchanger transfers the heat to a secondary circuit where a turbine is run to generate electricity.

Depending on the design, the fast neutrons used in the reactor can incubate additional fuel (breeding reaction) or potentially cause a reduction in long-lived waste materials through transmutation.

Like other fast reactors, the lead-cooled fast reactor can be used to incubate additional fuel or to convert long-lived waste material into shorter-lived or stable material by means of transmutation. The reactor core can be designed in such a way that the amount of heat generated per volume is relatively low. The lead alloy can dissipate all of the heat via an automatically adjusted circulation system; no primary circuit pumps are needed. The primary circuit also operates completely without pressure. In addition, lead has very good shielding properties against the ionising radiation emitted by the fuel.

One disadvantage of the system is that the lead-bismuth alloy must always be kept at temperatures above its melting point (min. 123 °C). If not, it will solidify and the entire reactor will become unusable. The coolant must also be filtered at great expense. Lead and bismuth have very high densities, so the system requires stronger structures due to the enormous weight. Bismuth is also very rare and expensive.

A research project on lead-cooled fast reactors was already underway in the USA in the 1940s, but was discontinued in 1950. In the Soviet Union, reactors of this type were developed to power submarines, and were used until 1996.

The 1990s/2000s witnessed a renewed interest in exploring the concept. Research and development projects are underway in the USA, China, Russia, South Korea and the EU, among others.

Problems that still remain unresolved include the minimisation of corrosion and erosion risks due to the liquid metal circulating in the primary circuit and the filtration of the coolant.