More than a dozen advanced reactor designs are in various stages of development. Some are evolutionary from the PWR, BWR and PHWR designs above, some are more radical departures. The former include the Advanced Boiling Water Reactor (ABWR), two of which are now operating with others under construction, and the planned passively safe ESBWR and AP1000 units (see Nuclear Power 2010 Program).

• The Integral Fast Reactor (IFR) was built, tested and evaluated during the 1980s and then retired under the Clinton administration in the 1990s due to nuclear non-proliferation policies of the administration. Recycling spent fuel is the core of its design and it therefore produces only a fraction of the waste of current reactors.
• The Pebble Bed Reactor, a High Temperature Gas Cooled Reactor (HTGCR), is designed so high temperatures reduce power output by doppler broadening of the fuel's neutron cross-section. It uses ceramic fuels so its safe operating temperatures exceed the power-reduction temperature range. Most designs are cooled by inert helium. Helium is not subject to steam explosions, resists neutron absorption leading to radioactivity, and does not dissolve contaminants that can become radioactive. Typical designs have more layers (up to 7) of passive containment than light water reactors (usually 3). A unique feature that may aid safety is that the fuel-balls actually form the core's mechanism, and are replaced one-by-one as they age. The design of the fuel makes fuel reprocessing expensive.
• The Small Sealed Transportable Autonomous Reactor (SSTAR) is being primarily researched and developed in the US, intended as a fast breeder reactor that is passively safe and could be remotely shut down in case the suspicion arises that it is being tampered with.
• The Clean And Environmentally Safe Advanced Reactor (CAESAR) is a nuclear reactor concept that uses steam as a moderator — this design is still in development.
• The Hydrogen Moderated Self-regulating Nuclear Power Module (HPM) is a reactor design emanating from the Los Alamos National Laboratory that uses uranium hydride as fuel.
• Subcritical reactors are designed to be safer and more stable, but pose a number of engineering and economic difficulties. One example is the Energy amplifier.
• Thorium based reactors. It is possible to convert Thorium-232 into U-233 in reactors specially designed for the purpose. In this way, thorium, which is more plentiful than uranium, can be used to breed U-233 nuclear fuel. U-233 is also believed to have favourable nuclear properties as compared to traditionally used U-235, including better neutron economy and lower production of long lived transuranic waste.
  • Advanced Heavy Water Reactor (AHWR)— A proposed heavy water moderated nuclear power reactor that will be the next generation design of the PHWR type. Under development in the Bhabha Atomic Research Centre (BARC), India.
  • KAMINI — A unique reactor using Uranium-233 isotope for fuel. Built in India by BARC and Indira Gandhi Center for Atomic Research (IGCAR).
  • India is also planning to build fast breeder reactors using the thorium – Uranium-233 fuel cycle. The FBTR (Fast Breeder Test Reactor) in operation at Kalpakkam (India) uses Plutonium as a fuel and liquid sodium as a coolant.

A generation 2 nuclear reactor is a design classification for a nuclear reactor, and refers to the class of commercial reactors built up to the end of the 1990s. Prototypical generation II reactors include the PWR, CANDU, BWR, AGR, and VVER.

These are contrasted to generation I reactors, which refer to the early prototype and power reactors, such as Shippingport, Magnox, Fermi 1, and Dresden. The nomenclature for reactor designs, describing four 'generations', was proposed by the US Department of Energy when it introduced the concept of generation IV reactors.

The designation generation II+ reactor is sometimes used for modernised generation II designs built post-2000, such as the Chinese CPR-1000, in competition with more expensive generation III reactor designs. Typically the modernisation includes improved safety systems and a 60 year design life.

Generation II reactor designs generally had an original design life of 30 or 40 years. However many generation II reactor are being life-extended to 50 or 60 years, and a second life-extension to 80 years may also be economic in many cases.

Generation 4 Nuclear Reactors (Gen IV) are a set of theoretical nuclear reactor designs currently being researched. Most of these designs are generally not expected to be available for commercial construction before 2030, with the exception of a version of the Very High Temperature Reactor (VHTR) called the Next Generation Nuclear Plant (NGNP). The NGNP is to be completed by 2021. Current reactors in operation around the world are generally considered second- or third-generation systems, with most of the first-generation systems having been retired some time ago. Research into these reactor types was officially started by the Generation IV International Forum (GIF) based on eight technology goals, including to improve nuclear safety, improve proliferation resistance, minimize waste and natural resource utilization, and decrease the cost to build and run such plants.

The reactors are intended for use in nuclear power plants to produce nuclear power from nuclear fuel.

Reactor types

Many reactor types were considered initially; however, the list was downsized to focus on the most promising technologies and those that could most likely meet the goals of the Gen IV initiative. Three systems are nominally thermal reactors and three are fast reactors. The VHTR is also being researched for potentially providing high quality process heat for hydrogen production. The fast reactors offer the possibility of burning actinides to further reduce waste and of being able to breed more fuel than they consume. These systems offer significant advances in sustainability, safety and reliability, economics, proliferation resistance and physical protection.

Thermal reactors

Very-high-temperature reactor (VHTR)

The very high temperature reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using helium or molten salt as the coolant. This reactor design envisions an outlet temperature of 1,000 °C. The reactor core can be either a prismatic-block or a pebble bed reactor design. The high temperatures enable applications such as process heat or hydrogen production via the thermochemical iodine-sulfur process. It would also be passively safe.

The planned construction of the first VHTR, the South African PBMR (pebble bed modular reactor), lost government funding in February, 2010. A pronounced increase of costs and concerns about possible unexpected technical problems had discouraged potential investors and customers.

Supercritical-water-cooled reactor (SCWR)

The supercritical water reactor (SCWR) is a concept that uses supercritical water as the working fluid. SCWRs are basically light water reactors (LWR) operating at higher pressure and temperatures with a direct, once-through cycle. As most commonly envisioned, it would operate on a direct cycle, much like a Boiling Water Reactor (BWR), but since it uses supercritical water (not to be confused with critical mass) as the working fluid, would have only one phase present, like the Pressurized Water Reactor (PWR). It could operate at much higher temperatures than both current PWRs and BWRs.

Supercritical water-cooled reactors (SCWRs) are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current LWRs) and considerable plant simplification.

The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil fuel fired boilers, a large number of which are also in use around the world. The SCWR concept is being investigated by 32 organizations in 13 countries.

Molten-salt reactor (MSR)

A molten salt reactor is a type of nuclear reactor where the coolant is a molten salt. There have been many designs put forward for this type of reactor and a few prototypes built. The early concepts and many current ones rely on nuclear fuel dissolved in the molten fluoride salt as uranium tetrafluoride (UF₄) or thorium tetrafluoride (ThF₄), the fluid would reach criticality by flowing into a graphite core which would also serve as the moderator. Many current concepts rely on fuel that is dispersed in a graphite matrix with the molten salt providing low pressure, high temperature cooling.

Fast reactors

Gas-cooled fast reactor (GFR)

The gas-cooled fast reactor (GFR) system features a fast-neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reactor is helium-cooled, with an outlet temperature of 850 °C and using a direct Brayton cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being considered based on pin- or plate-based fuel assemblies or prismatic blocks.

Sodium-cooled fast reactor (SFR)

The SFR is a project that builds on two closely related existing projects, the liquid metal fast breeder reactor and the Integral Fast Reactor.

The goals are to increase the efficiency of uranium usage by breeding plutonium and eliminating the need for transuranic isotopes ever to leave the site. The reactor design uses an unmoderated core running on fast neutrons, designed to allow any transuranic isotope to be consumed (and in some cases used as fuel). In addition to the benefits of removing the long half-life transuranics from the waste cycle, the SFR fuel expands when the reactor overheats, and the chain reaction automatically slows down. In this manner, it is passively safe.

The Integral Fast Reactor or IFR is a design for a nuclear reactor with a specialized nuclear fuel cycle. A prototype of the reactor was built, but the project was cancelled before it could be copied elsewhere.

The SFR reactor concept is cooled by liquid sodium and fueled by a metallic alloy of uranium and plutonium. The fuel is contained in steel cladding with liquid sodium filling in the space between the clad elements which make up the fuel assembly. One of the design challenges of an SFR is the risks of handling sodium, which reacts explosively if it comes into contact with water. However, the use of liquid metal instead of water as coolant allows the system to work at atmospheric pressure, reducing the risk of leakage.

Lead-cooled fast reactor (LFR)

The lead-cooled fast reactor features a fast-neutron-spectrum lead or lead/bismuth eutectic (LBE) liquid-metal-cooled reactor with a closed fuel cycle. Options include a range of plant ratings, including a "battery" of 50 to 150 MW of electricity that features a very long refueling interval, a modular system rated at 300 to 400 MW, and a large monolithic plant option at 1,200 MW. (The term battery refers to the long-life, factory-fabricated core, not to any provision for electrochemical energy conversion.) The fuel is metal or nitride-based containing fertile uranium and transuranics. The LFR is cooled by natural convection with a reactor outlet coolant temperature of 550 °C, possibly ranging up to 800 °C with advanced materials. The higher temperature enables the production of hydrogen by thermochemical processes.

A generation III nuclear reactor is a development of any of the generation II nuclear reactor designs incorporating evolutionary improvements in design developed during the lifetime of the generation II reactor designs. These include improved fuel technology, superior thermal efficiency, passive safety systems and standardized design for reduced maintenance and capital costs.

Improvements in reactor technology result in a longer operational life (60 years of operation, extendable to 120+ years of operation prior to complete overhaul and reactor pressure vessel replacement) compared with currently used generation II reactors (designed for 40 years of operation, extendable to 80+ years of operation prior to complete overhaul and RPV replacement). Furthermore, core damage frequencies for these reactors are lower than for Generation II reactors — 60 core damage events per 1000 million reactor–year for the EPR; 3 core damage events per 1000 million reactor–year for the ESBWR significantly lower than the 10,000 core damage events per 1000 million reactor–year for BWR/4 generation II reactors.

The first generation III reactors were built in Japan, while several others have been approved for construction in Europe. A Westinghouse AP1000 reactor is scheduled to become operational in Sanmen, China in 2013.

Generation III reactors

• Advanced Boiling Water Reactor (ABWR) — A GE design that first went online in Japan in 1996.
• Advanced Pressurized Water Reactor (APWR) — developed by Mitsubishi Heavy Industries.
• Enhanced CANDU 6 (EC6) — developed by Atomic Energy of Canada Limited.
• VVER-1000/392 (PWR) — in various modifications into AES-91 and AES-92

Designs not adopted

• AP600 — A Westinghouse Electric Company design that received final design approval from the NRC in 1998; the EIA states that "Westinghouse has deemphasized the AP600 in favor of the larger, though potentially even less expensive (on a cost per kilowatt or capacity basis) AP1000 design."
• System 80+ — a Combustion Engineering (now incorporated into Westinghouse) design, which "provides a basis for the APR1400 (Generation III+) design that has been developed in Korea for future deployment and possible export."

Generation III+ reactors

Generation III+ designs offer significant improvements in safety and economics over Generation III advanced reactor designs certified by the NRC in the 1990s.

• Advanced CANDU Reactor (ACR-1000)
• AP1000 — based on the AP600 with increased power output
• European Pressurized Reactor (EPR) — an evolutionary descendant of the Framatome N4 and Siemens Power Generation Division KONVOI reactors.
• Economic Simplified Boiling Water Reactor (ESBWR) — based on the ABWR
• APR-1400 — an advanced PWR design evolved from the U.S. System 80+, which is the basis for the Korean Next Generation Reactor or KNGR
• VVER-1200/392M (PWR) — in design of AES-2006 with mainly passive safety features
• VVER-1200/491 (PWR) — in design of AES-2006 with mainly active safety features, international sold as MIR.1200
• EU-ABWR — based on the ABWR with increased powert output and compliance with EU safety standard.
• Advanced PWR (APWR) — 4th Generation of PWR from Mitsubishi Heavy Industries

Generation III++ reactors

• B&W mPower — an Advanced Light Water Reactor in development by Babcock and Wilcox and Bechtel

Nuclear reactor coolant is coolant in a nuclear reactor that is used to remove heat from the nuclear reactor core and transfer it to electrical generators and the environment. Frequently a chain of two coolant loops is used because the primary coolant loop takes on short-term radioactivity from the reactor.

Almost all currently operating nuclear power plants are light water reactors using ordinary water under high pressure as coolant and neutron moderator. About 1/3 are boiling water reactors where the primary coolant undergoes phase change to steam inside the reactor. About 2/3 are pressurized water reactors at even higher pressure. Current reactors stay under the critical point at around 374 °C and 218 bar where the distinction between liquid and gas disappears, which limits thermal efficiency, but the proposed supercritical water reactor would operate above this point.

Fast reactors have a high power density and do not need neutron moderation. Most have been liquid metal cooled reactors using molten sodium. Lead and other metals have also been proposed and occasionally used.

Molten salts share with metals the advantage of low vapor pressure even at high temperatures, and are less chemically reactive than sodium. Salts containing light elements like FLiBe can also provide moderation. In the Molten-Salt Reactor Experiment it even served as a solvent carrying the nuclear fuel.

Gases have also been used as coolant. Helium is extremely inert both chemically and with respect to nuclear reactions but has a low heat capacity, necessitating rapid circulation. Carbon dioxide has also been used. Gases of course need to be under pressure for sufficient density at high temperature.

Nuclear reactor coolants

Coolant	Melting point	Boiling point
Light water at 155 bar		345 °C
Mercury	-38.83 °C	356.73 °C
NaK eutectic	-11 °C	785 °C
Sodium	97.72 °C	883 °C
FLiBe	459 °C	1430 °C
Lead	327.46 °C	1749 °C
Lead-bismuth eutectic	123.5 °C	1670 °C

Just as many conventional thermal power stations generate electricity by harnessing the thermal energy released from burning fossil fuels, nuclear power plants convert the energy released from the nucleus of an atom, typically via nuclear fission.

When a relatively large fissile atomic nucleus (usually uranium-235 or plutonium-239) absorbs a neutron, a fission of the atom often results. Fission splits the atom into two or more smaller nuclei with kinetic energy (known as fission products) and also releases gamma radiation and free neutrons. A portion of these neutrons may later be absorbed by other fissile atoms and create more fissions, which release more neutrons, and so on.

This nuclear chain reaction can be controlled by using neutron poisons and neutron moderators to change the portion of neutrons that will go on to cause more fissions. Nuclear reactors generally have automatic and manual systems to shut the fission reaction down if unsafe conditions are detected.

A cooling system removes heat from the reactor core and transports it to another area of the plant, where the thermal energy can be harnessed to produce electricity or to do other useful work. Typically the hot coolant will be used as a heat source for a boiler, and the pressurized steam from that boiler will power one or more steam turbine driven electrical generators.

There are many different reactor designs, utilizing different fuels and coolants and incorporating different control schemes. Some of these designs have been engineered to meet a specific need. Reactors for nuclear submarines and large naval ships, for example, commonly use highly enriched uranium as a fuel. This fuel choice increases the reactor's power density and extends the usable life of the nuclear fuel load, but is more expensive and a greater risk to nuclear proliferation than some of the other nuclear fuels.

A number of new designs for nuclear power generation, collectively known as the Generation IV reactors, are the subject of active research and may be used for practical power generation in the future. Many of these new designs specifically attempt to make fission reactors cleaner, safer and/or less of a risk to the proliferation of nuclear weapons. Passively safe plants (such as the ESBWR) are available to be built and other designs that are believed to be nearly fool-proof are being pursued. Fusion reactors, which may be viable in the future, diminish or eliminate many of the risks associated with nuclear fission.

There are two types of nuclear power in current use:

• The Radioisotope thermoelectric generator produces heat through passive radioactive decay. Some radioisotope thermoelectric generators have been created to power space probes (for example, the Cassini probe), some lighthouses in the former Soviet Union, and some pacemakers. The heat output of these generators diminishes with time; the heat is converted to electricity utilising the thermoelectric effect.
• Nuclear fission reactors produce heat through a controlled nuclear chain reaction in a critical mass of fissile material. All current nuclear power plants are critical fission reactors, which are the focus of this article. The output of fission reactors is controllable. There are several subtypes of critical fission reactors, which can be classified as Generation I, Generation II and Generation III. All reactors will be compared to the Pressurized Water Reactor (PWR), as that is the standard modern reactor design.

Pressurized Water Reactors (PWR)

These reactors use a pressure vessel to contain the nuclear fuel, control rods, moderator, and coolant. They are cooled and moderated by high pressure liquid water. The hot radioactive water that leaves the pressure vessel is looped through a steam generator, which in turn heats a secondary (non-radioactive) loop of water to steam that can run turbines. They are the majority of current reactors, and are generally considered the safest and most reliable technology currently in large scale deployment. This is a thermal neutron reactor design, the newest of which are the VVER-1200, Advanced Pressurized Water Reactor and the European Pressurized Reactor. United States Naval reactors are of this type.

Boiling Water Reactors (BWR)

A BWR is like a PWR without the steam generator. A boiling water reactor is cooled and moderated by water like a PWR, but at a lower pressure, which allows the water to boil inside the pressure vessel producing the steam that runs the turbines. Unlike a PWR, there is no primary and secondary loop. The thermal efficiency of these reactors can be higher, and they can be simpler, and even potentially more stable and safe. This is a thermal neutron reactor design, the newest of which are the Advanced Boiling Water Reactor and the Economic Simplified Boiling Water Reactor.

Pressurized Heavy Water Reactor (PHWR)

A Canadian design (known as CANDU), these reactors are heavy-water-cooled and -moderated Pressurized-Water reactors. Instead of using a single large pressure vessel as in a PWR, the fuel is contained in hundreds of pressure tubes. These reactors are fueled with natural uranium and are thermal neutron reactor designs. PHWRs can be refueled while at full power, which makes them very efficient in their use of uranium (it allows for precise flux control in the core). CANDU PHWRs have been built in Canada, Argentina, China, India (pre-NPT), Pakistan (pre-NPT), Romania, and South Korea. India also operates a number of PHWRs, often termed 'CANDU-derivatives', built after the Government of Canada halted nuclear dealings with India following the 1974 Smiling Buddha nuclear weapon test.

Reaktor Bolshoy Moschnosti Kanalniy (High Power Channel Reactor) (RBMK)

A Soviet design, built to produce plutonium as well as power. RBMKs are water cooled with a graphite moderator. RBMKs are in some respects similar to CANDU in that they are refuelable during power operation and employ a pressure tube design instead of a PWR-style pressure vessel. However, unlike CANDU they are very unstable and large, making containment buildings for them expensive. A series of critical safety flaws have also been identified with the RBMK design, though some of these were corrected following the Chernobyl accident. Their main attraction is their use of light water and un-enriched uranium. As of 2010, 11 remain open, mostly due to safety improvements and help from international safety agencies such as the DOE. Despite these safety improvements, RBMK reactors are still considered one of the most dangerous reactor designs in use. RBMK reactors were deployed only in the former Soviet Union.

Gas Cooled Reactor (GCR) and Advanced Gas Cooled Reactor (AGR)

These are generally graphite moderated and CO₂ cooled. They can have a high thermal efficiency compared with PWRs due to higher operating temperatures. There are a number of operating reactors of this design, mostly in the United Kingdom, where the concept was developed. Older designs (i.e. Magnox stations) are either shut down or will be in the near future. However, the AGCRs have an anticipated life of a further 10 to 20 years. This is a thermal neutron reactor design. Decommissioning costs can be high due to large volume of reactor core.

Liquid Metal Fast Breeder Reactor (LMFBR)

This is a reactor design that is cooled by liquid metal, totally unmoderated, and produces more fuel than it consumes. They are said to "breed" fuel, because they produce fissionable fuel during operation because of neutron capture. These reactors can function much like a PWR in terms of efficiency, and do not require much high pressure containment, as the liquid metal does not need to be kept at high pressure, even at very high temperatures. BN-350 and BN-600 in USSR and Superphénix in France were a reactor of this type, as was Fermi-I in the United States. The Monju reactor in Japan suffered a sodium leak in 1995 and is pending restart earliest in February 2010. All of them use/used liquid sodium. These reactors are fast neutron, not thermal neutron designs. These reactors come in two types:

Lead cooled

Using lead as the liquid metal provides excellent radiation shielding, and allows for operation at very high temperatures. Also, lead is (mostly) transparent to neutrons, so fewer neutrons are lost in the coolant, and the coolant does not become radioactive. Unlike sodium, lead is mostly inert, so there is less risk of explosion or accident, but such large quantities of lead may be problematic from toxicology and disposal points of view. Often a reactor of this type would use a lead-bismuth eutectic mixture. In this case, the bismuth would present some minor radiation problems, as it is not quite as transparent to neutrons, and can be transmuted to a radioactive isotope more readily than lead. The Russian Alfa class submarine uses a lead-bismuth-cooled fast reactor as its main power plant.

Sodium cooled

Most LMFBRs are of this type. The sodium is relatively easy to obtain and work with, and it also manages to actually prevent corrosion on the various reactor parts immersed in it. However, sodium explodes violently when exposed to water, so care must be taken, but such explosions wouldn't be vastly more violent than (for example) a leak of superheated fluid from a SCWR or PWR. EBR-I, the first reactor to have a core meltdown, was of this type.

Pebble Bed Reactors (PBR)

These use fuel molded into ceramic balls, and then circulate gas through the balls. The result is an efficient, low-maintenance, very safe reactor with inexpensive, standardized fuel. The prototype was the AVR.

Molten Salt Reactors

These dissolve the fuels in fluoride salts, or use fluoride salts for coolant. These have many safety features, high efficiency and a high power density suitable for vehicles. Notably, they have no high pressures or flammable components in the core. The prototype was the MSRE, which also used Thorium's fuel cycle to produce 0.1% of the radioactive waste of standard reactors.

Aqueous Homogeneous Reactor (AHR)

These reactors use soluble nuclear salts dissolved in water and mixed with a coolant and a neutron moderator.