Nuclear reactor

When a large fissile atomic nucleus such as uranium-235 or plutonium-239 absorbs a neutron, it may undergo nuclear fission. The heavy nucleus splits into two or more lighter nuclei, (the fission products), releasing kinetic energy, gamma radiation, and free neutrons. A portion of these neutrons may be absorbed by other fissile atoms and trigger further fission events, which release more neutrons, and so on. This is known as a nuclear chain reaction.

To control such a nuclear chain reaction, Control rods containing neutron poisons and neutron moderators can change the portion of neutrons that will go on to cause more fission.[4] Nuclear reactors generally have automatic and manual systems to shut the fission reaction down if monitoring detects unsafe conditions.[5]

The reactor core generates heat in a number of ways:

• The kinetic energy of fission products is converted to thermal energy when these nuclei collide with nearby atoms.
• The reactor absorbs some of the gamma rays produced during fission and converts their energy into heat.
• Heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption. This decay heat source will remain for some time even after the reactor is shut down.

A kilogram of uranium-235 (U-235) converted via nuclear processes releases approximately three million times more energy than a kilogram of coal burned conventionally (7.2 × 10¹³ joules per kilogram of uranium-235 versus 2.4 × 10⁷ joules per kilogram of coal).[6][7]

A nuclear reactor coolant — usually water but sometimes a gas or a liquid metal (like liquid sodium or lead) or molten salt — is circulated past the reactor core to absorb the heat that it generates. The heat is carried away from the reactor and is then used to generate steam. Most reactor systems employ a cooling system that is physically separated from the water that will be boiled to produce pressurized steam for the turbines, like the pressurized water reactor. However, in some reactors the water for the steam turbines is boiled directly by the reactor core; for example the boiling water reactor.[8]

The rate of fission reactions within a reactor core can be adjusted by controlling the quantity of neutrons that are able to induce further fission events. Nuclear reactors typically employ several methods of neutron control to adjust the reactor's power output. Some of these methods arise naturally from the physics of radioactive decay and are simply accounted for during the reactor's operation, while others are mechanisms engineered into the reactor design for a distinct purpose.

The fastest method for adjusting levels of fission-inducing neutrons in a reactor is via movement of the control rods. Control rods are made of neutron poisons and therefore absorb neutrons. When a control rod is inserted deeper into the reactor, it absorbs more neutrons than the material it displaces—often the moderator. This action results in fewer neutrons available to cause fission and reduces the reactor's power output. Conversely, extracting the control rod will result in an increase in the rate of fission events and an increase in power.

The physics of radioactive decay also affects neutron populations in a reactor. One such process is delayed neutron emission by a number of neutron-rich fission isotopes. These delayed neutrons account for about 0.65% of the total neutrons produced in fission, with the remainder (termed "prompt neutrons") released immediately upon fission. The fission products which produce delayed neutrons have half-lives for their decay by neutron emission that range from milliseconds to as long as several minutes, and so considerable time is required to determine exactly when a reactor reaches the critical point. Keeping the reactor in the zone of chain reactivity where delayed neutrons are necessary to achieve a critical mass state allows mechanical devices or human operators to control a chain reaction in "real time"; otherwise the time between achievement of criticality and nuclear meltdown as a result of an exponential power surge from the normal nuclear chain reaction, would be too short to allow for intervention. This last stage, where delayed neutrons are no longer required to maintain criticality, is known as the prompt critical point. There is a scale for describing criticality in numerical form, in which bare criticality is known as zero dollars and the prompt critical point is one dollar, and other points in the process interpolated in cents.

In some reactors, the coolant also acts as a neutron moderator. A moderator increases the power of the reactor by causing the fast neutrons that are released from fission to lose energy and become thermal neutrons. Thermal neutrons are more likely than fast neutrons to cause fission. If the coolant is a moderator, then temperature changes can affect the density of the coolant/moderator and therefore change power output. A higher temperature coolant would be less dense, and therefore a less effective moderator.

In other reactors the coolant acts as a poison by absorbing neutrons in the same way that the control rods do. In these reactors power output can be increased by heating the coolant, which makes it a less dense poison. Nuclear reactors generally have automatic and manual systems to scram the reactor in an emergency shut down. These systems insert large amounts of poison (often boron in the form of boric acid) into the reactor to shut the fission reaction down if unsafe conditions are detected or anticipated.[9]

Most types of reactors are sensitive to a process variously known as xenon poisoning, or the iodine pit. The common fission product Xenon-135 produced in the fission process acts as a neutron poison that absorbs neutrons and therefore tends to shut the reactor down. Xenon-135 accumulation can be controlled by keeping power levels high enough to destroy it by neutron absorption as fast as it is produced. Fission also produces iodine-135, which in turn decays (with a half-life of 6.57 hours) to new xenon-135. When the reactor is shut down, iodine-135 continues to decay to xenon-135, making restarting the reactor more difficult for a day or two, as the xenon-135 decays into cesium-135, which is not nearly as poisonous as xenon-135, with a half-life of 9.2 hours. This temporary state is the "iodine pit." If the reactor has sufficient extra reactivity capacity, it can be restarted. As the extra xenon-135 is transmuted to xenon-136, which is much less a neutron poison, within a few hours the reactor experiences a "xenon burnoff (power) transient". Control rods must be further inserted to replace the neutron absorption of the lost xenon-135. Failure to properly follow such a procedure was a key step in the Chernobyl disaster.[10]

Reactors used in nuclear marine propulsion (especially nuclear submarines) often cannot be run at continuous power around the clock in the same way that land-based power reactors are normally run, and in addition often need to have a very long core life without refueling. For this reason many designs use highly enriched uranium but incorporate burnable neutron poison in the fuel rods.[11] This allows the reactor to be constructed with an excess of fissionable material, which is nevertheless made relatively safe early in the reactor's fuel burn cycle by the presence of the neutron-absorbing material which is later replaced by normally produced long-lived neutron poisons (far longer-lived than xenon-135) which gradually accumulate over the fuel load's operating life.

The energy released in the fission process generates heat, some of which can be converted into usable energy. A common method of harnessing this thermal energy is to use it to boil water to produce pressurized steam which will then drive a steam turbine that turns an alternator and generates electricity.[9]

All commercial power reactors are based on nuclear fission. They generally use uranium and its product plutonium as nuclear fuel, though a thorium fuel cycle is also possible. Fission reactors can be divided roughly into two classes, depending on the energy of the neutrons that sustain the fission chain reaction:

• Thermal neutron reactors (the most common type of nuclear reactor) use slowed or thermal neutrons to keep up the fission of their fuel. Almost all current reactors are of this type. These contain neutron moderator materials that slow neutrons until their neutron temperature is thermalized, that is, until their kinetic energy approaches the average kinetic energy of the surrounding particles. Thermal neutrons have a far higher cross section (probability) of fissioning the fissile nuclei uranium-235, plutonium-239, and plutonium-241, and a relatively lower probability of neutron capture by uranium-238 (U-238) compared to the faster neutrons that originally result from fission, allowing use of low-enriched uranium or even natural uranium fuel. The moderator is often also the coolant, usually water under high pressure to increase the boiling point. These are surrounded by a reactor vessel, instrumentation to monitor and control the reactor, radiation shielding, and a containment building.
• Fast neutron reactors use fast neutrons to cause fission in their fuel. They do not have a neutron moderator, and use less-moderating coolants. Maintaining a chain reaction requires the fuel to be more highly enriched in fissile material (about 20% or more) due to the relatively lower probability of fission versus capture by U-238. Fast reactors have the potential to produce less transuranic waste because all actinides are fissionable with fast neutrons,[23] but they are more difficult to build and more expensive to operate. Overall, fast reactors are less common than thermal reactors in most applications. Some early power stations were fast reactors, as are some Russian naval propulsion units. Construction of prototypes is continuing (see fast breeder or generation IV reactors).

In principle, fusion power could be produced by nuclear fusion of elements such as the deuterium isotope of hydrogen. While an ongoing rich research topic since at least the 1940s, no self-sustaining fusion reactor for power generation has ever been built.

Used by thermal reactors:

• Graphite-moderated reactors
• Water moderated reactors
  • Heavy-water reactors (Used in Canada,[24] India, Argentina, China, Pakistan, Romania and South Korea).[25]
  • Light-water-moderated reactors (LWRs). Light-water reactors (the most common type of thermal reactor) use ordinary water to moderate and cool the reactors.[24] When at operating temperature, if the temperature of the water increases, its density drops, and fewer neutrons passing through it are slowed enough to trigger further reactions. That negative feedback stabilizes the reaction rate. Graphite and heavy-water reactors tend to be more thoroughly thermalized than light water reactors. Due to the extra thermalization, these types can use natural uranium/unenriched fuel.
• Light-element-moderated reactors.
  • Molten salt reactors (MSRs) are moderated by light elements such as lithium or beryllium, which are constituents of the coolant/fuel matrix salts LiF and BeF₂.
  • Liquid metal cooled reactors, such as those whose coolant is a mixture of lead and bismuth, may use BeO as a moderator.
• Organically moderated reactors (OMR) use biphenyl and terphenyl as moderator and coolant.

Treatment of the interior part of a VVER-1000 reactor frame at Atommash.

In thermal nuclear reactors (LWRs in specific), the coolant acts as a moderator that must slow down the neutrons before they can be efficiently absorbed by the fuel.

• Water cooled reactor. These constitute the great majority of operational nuclear reactors: as of 2014, 93% of the world's nuclear reactors are water cooled, providing about 95% of the world's total nuclear generation capacity.[22]
  • Pressurized water reactor (PWR) Pressurized water reactors constitute the large majority of all Western nuclear power plants.
    • A primary characteristic of PWRs is a pressurizer, a specialized pressure vessel. Most commercial PWRs and naval reactors use pressurizers. During normal operation, a pressurizer is partially filled with water, and a steam bubble is maintained above it by heating the water with submerged heaters. During normal operation, the pressurizer is connected to the primary reactor pressure vessel (RPV) and the pressurizer "bubble" provides an expansion space for changes in water volume in the reactor. This arrangement also provides a means of pressure control for the reactor by increasing or decreasing the steam pressure in the pressurizer using the pressurizer heaters.
    • Pressurized heavy water reactors are a subset of pressurized water reactors, sharing the use of a pressurized, isolated heat transport loop, but using heavy water as coolant and moderator for the greater neutron economies it offers.
  • Boiling water reactor (BWR)
    • BWRs are characterized by boiling water around the fuel rods in the lower portion of a primary reactor pressure vessel. A boiling water reactor uses ²³⁵U, enriched as uranium dioxide, as its fuel. The fuel is assembled into rods housed in a steel vessel that is submerged in water. The nuclear fission causes the water to boil, generating steam. This steam flows through pipes into turbines. The turbines are driven by the steam, and this process generates electricity.[26] During normal operation, pressure is controlled by the amount of steam flowing from the reactor pressure vessel to the turbine.
  • Supercritical water reactor (SCWR)
    • SCWRs are a Generation IV reactor concept where the reactor is operated at supercritical pressures and water is heated to a supercritical fluid, which never undergoes a transition to steam yet behaves like saturated steam, to power a steam generator.
  • Pool-type reactor can refer to open pool reactors which are water cooled, but not to be confused with pool type LMFBRs which are sodium cooled
  • Some reactors have been cooled by heavy water which also served as a moderator. Examples include:
    • Early CANDU reactors (later ones use heavy water moderator but light water coolant)
    • DIDO class research reactors
• Liquid metal cooled reactor. Since water is a moderator, it cannot be used as a coolant in a fast reactor. Liquid metal coolants have included sodium, NaK, lead, lead-bismuth eutectic, and in early reactors, mercury.
  • Sodium-cooled fast reactor
  • Lead-cooled fast reactor
• Gas cooled reactors are cooled by a circulating gas. In commercial nuclear power plants carbon dioxide has usually been used, for example in current British AGR nuclear power plants and formerly in a number of first generation British, French, Italian, & Japanese plants. Nitrogen[27] and helium have also been used, helium being considered particularly suitable for high temperature designs. Utilization of the heat varies, depending on the reactor. Commercial nuclear power plants run the gas through a heat exchanger to make steam for a steam turbine. Some experimental designs run hot enough that the gas can directly power a gas turbine.
• Molten salt reactors (MSRs) are cooled by circulating a molten salt, typically a eutectic mixture of fluoride salts, such as FLiBe. In a typical MSR, the coolant is also used as a matrix in which the fissile material is dissolved.
• Organic nuclear reactors use organic fluids such as biphenyl and terphenyl as coolant rather than water.

• Generation I reactor (early prototypes such as Shippingport Atomic Power Station, research reactors, non-commercial power producing reactors)
• Generation II reactor (most current nuclear power plants, 1965–1996)
• Generation III reactor (evolutionary improvements of existing designs, 1996–present)
• Generation III+ reactor (evolutionary development of Gen III reactors, offering improvements in safety over Gen III reactor designs, 2017–present)[28]
• Generation IV reactor (technologies still under development; unknown start date, possibly 2030)[29]
• Generation V reactors: Reactors that are purely theoretical and are not the subject of intense research.

In 2003, the French Commissariat à l'Énergie Atomique (CEA) was the first to refer to "Gen II" types in Nucleonics Week.[30]

The first mention of "Gen III" was in 2000, in conjunction with the launch of the Generation IV International Forum (GIF) plans.

"Gen IV" was named in 2000, by the United States Department of Energy (DOE), for developing new plant types.[31]

• Solid fueled
• Fluid fueled
  • Aqueous homogeneous reactor
  • Molten salt reactor
• Gas fueled (theoretical)

• Cubical
• Cylindrical
• Octagonal
• Spherical
• Slab
• Annulus

• Electricity
  • Nuclear power plants including small modular reactors
• Propulsion, see nuclear propulsion
  • Nuclear marine propulsion
  • Various proposed forms of rocket propulsion
• Other uses of heat
  • Desalination
  • Heat for domestic and industrial heating
  • Hydrogen production for use in a hydrogen economy
• Production reactors for transmutation of elements
  • Breeder reactors are capable of producing more fissile material than they consume during the fission chain reaction (by converting fertile U-238 to Pu-239, or Th-232 to U-233). Thus, a uranium breeder reactor, once running, can be refueled with natural or even depleted uranium, and a thorium breeder reactor can be refueled with thorium; however, an initial stock of fissile material is required.[32]
  • Creating various radioactive isotopes, such as americium for use in smoke detectors, and cobalt-60, molybdenum-99 and others, used for imaging and medical treatment.
  • Production of materials for nuclear weapons such as weapons-grade plutonium
• Providing a source of neutron radiation (for example with the pulsed Godiva device) and positron radiation (e.g. neutron activation analysis and potassium-argon dating)
• Research reactor: Typically reactors used for research and training, materials testing, or the production of radioisotopes for medicine and industry. These are much smaller than power reactors or those propelling ships, and many are on university campuses. There are about 280 such reactors operating, in 56 countries. Some operate with high-enriched uranium fuel, and international efforts are underway to substitute low-enriched fuel.[33]

Diablo Canyon — a PWR

• Pressurized water reactors (PWR) [moderator: high-pressure water; coolant: high-pressure water]

These reactors use a pressure vessel to contain the nuclear fuel, control rods, moderator, and coolant. The hot radioactive water that leaves the pressure vessel is looped through a steam generator, which in turn heats a secondary (nonradioactive) loop of water to steam that can run turbines. They represent the majority (around 80%) of current reactors. This is a thermal neutron reactor design, the newest of which are the Russian VVER-1200, Japanese Advanced Pressurized Water Reactor, American AP1000, Chinese Hualong Pressurized Reactor and the Franco-German European Pressurized Reactor. All the United States Naval reactors are of this type.

• Boiling water reactors (BWR) [moderator: low-pressure water; coolant: low-pressure water]

A BWR is like a PWR without the steam generator. The lower pressure of its cooling water allows it 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.

The CANDU Qinshan Nuclear Power Plant

• Pressurized Heavy Water Reactor (PHWR) [moderator: high-pressure heavy water; coolant: high-pressure heavy water]

A Canadian design (known as CANDU), very similar to PWRs but using heavy water. While heavy water is significantly more expensive than ordinary water, it has greater neutron economy (creates a higher number of thermal neutrons), allowing the reactor to operate without fuel enrichment facilities. 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, Pakistan, 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.

The Ignalina Nuclear Power Plant — a RBMK type (closed 2009)

• Reaktor Bolshoy Moschnosti Kanalniy (High Power Channel Reactor) (RBMK) [moderator: graphite; coolant: high-pressure water]

A Soviet design, 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 disaster. Their main attraction is their use of light water and unenriched uranium. As of 2021, 9 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.

The Magnox Sizewell A nuclear power station

The Torness nuclear power station — an AGR

• Gas-cooled reactor (GCR) and advanced gas-cooled reactor (AGR) [moderator: graphite; coolant: carbon dioxide]

These designs an 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 AGRs 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) [moderator: none; coolant: liquid metal]

Scaled – down model of TOPAZ nuclear reactor

This totally unmoderated reactor design 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. These reactors are fast neutron, not thermal neutron designs. These reactors come in two types:

The Superphénix, closed in 1998, was one of the few FBRs

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 TOPAZ, BN-350 and BN-600 in USSR; Superphénix in France; and Fermi-I in the United States were reactors 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 would not be more violent than (for example) a leak of superheated fluid from a pressurized-water reactor. The Monju reactor in Japan suffered a sodium leak in 1995 and could not be restarted until May 2010. The EBR-I, the first reactor to have a core meltdown, in 1955, was also a sodium-cooled reactor.

• Pebble-bed reactors (PBR) [moderator: graphite; coolant: helium]

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 and the HTR-10 is operating in China, where the HTR-PM is being developed. The HTR-PM is expected to be the first generation IV reactor to enter operation.[34]

• Molten salt reactors (MSR) [moderator: graphite; coolant: molten salt mixture]

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 the Thorium fuel cycle. As a breeder reactor type, it reprocesses the spent fuel, extracting both Uranium and transuranics, leaving only 0.1% of transuranic waste compared to conventional once-through uranium-fueled light water reactors currently in use. A separate issue is the radioactive fission products, which are not reprocessable and need to be disposed of as with conventional reactors.

• Aqueous Homogeneous Reactor (AHR) [moderator: high-pressure light or heavy water; coolant: high-pressure light or heavy water]

These reactors use as fuel soluble nuclear salts (usually uranium sulfate or uranium nitrate) dissolved in water and mixed with the coolant and the moderator. As of April 2006, only five AHRs were in operation.[35]

More than a dozen advanced reactor designs are in various stages of development.[36] 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 Economic Simplified Boiling Water Reactor (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.[37]
• 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 Reduced moderation water reactor builds upon the Advanced boiling water reactor(ABWR) that is presently in use, it is not a complete fast reactor instead using mostly epithermal neutrons, which are between thermal and fast neutrons in speed.
• 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 four times more abundant than uranium, can be used to breed U-233 nuclear fuel.[38] 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.
  • China, which has control of the Cerro Impacto deposit, has a reactor and hopes to replace coal energy with nuclear energy.[39]

Rolls-Royce aims to sell nuclear reactors for the production of synfuel for aircraft.[40]

Generation IV reactors are a set of theoretical nuclear reactor designs currently being researched. These designs are generally not expected to be available for commercial construction before 2030. Current reactors in operation around the world are generally considered second- or third-generation systems, with 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. The primary goals being to improve nuclear safety, improve proliferation resistance, minimize waste and natural resource utilization, and to decrease the cost to build and run such plants.[41]

• Gas-cooled fast reactor
• Lead-cooled fast reactor
• Molten salt reactor
• Sodium-cooled fast reactor
• Supercritical water reactor
• Very-high-temperature reactor

Generation V reactors are designs which are theoretically possible, but which are not being actively considered or researched at present. Though some generation V reactors could potentially be built with current or near term technology, they trigger little interest for reasons of economics, practicality, or safety.

• Liquid-core reactor. A closed loop liquid-core nuclear reactor, where the fissile material is molten uranium or uranium solution cooled by a working gas pumped in through holes in the base of the containment vessel.
• Gas-core reactor. A closed loop version of the nuclear lightbulb rocket, where the fissile material is gaseous uranium hexafluoride contained in a fused silica vessel. A working gas (such as hydrogen) would flow around this vessel and absorb the UV light produced by the reaction. This reactor design could also function as a rocket engine, as featured in Harry Harrison's 1976 science-fiction novel Skyfall. In theory, using UF₆ as a working fuel directly (rather than as a stage to one, as is done now) would mean lower processing costs, and very small reactors. In practice, running a reactor at such high power densities would probably produce unmanageable neutron flux, weakening most reactor materials, and therefore as the flux would be similar to that expected in fusion reactors, it would require similar materials to those selected by the International Fusion Materials Irradiation Facility.
  • Gas core EM reactor. As in the gas core reactor, but with photovoltaic arrays converting the UV light directly to electricity.[42] This approach is similar to the experimentally proved photoelectric effect that would convert the X-rays generated from aneutronic fusion into electricity, by passing the high energy photons through an array of conducting foils to transfer some of their energy to electrons, the energy of the photon is captured electrostatically, similar to a capacitor. Since X-rays can go through far greater material thickness than electrons, many hundreds or thousands of layers are needed to absorb the X-rays.[43]
• Fission fragment reactor. A fission fragment reactor is a nuclear reactor that generates electricity by decelerating an ion beam of fission byproducts instead of using nuclear reactions to generate heat. By doing so, it bypasses the Carnot cycle and can achieve efficiencies of up to 90% instead of 40–45% attainable by efficient turbine-driven thermal reactors. The fission fragment ion beam would be passed through a magnetohydrodynamic generator to produce electricity.
• Hybrid nuclear fusion. Would use the neutrons emitted by fusion to fission a blanket of fertile material, like U-238 or Th-232 and transmute other reactor's spent nuclear fuel/nuclear waste into relatively more benign isotopes.

Controlled nuclear fusion could in principle be used in fusion power plants to produce power without the complexities of handling actinides, but significant scientific and technical obstacles remain. Several fusion reactors have been built, but only recently reactors have been able to release more energy than the amount of energy used in the process. Despite research having started in the 1950s, no commercial fusion reactor is expected before 2050. The ITER project is currently leading the effort to harness fusion power.

The amount of energy in the reservoir of nuclear fuel is frequently expressed in terms of "full-power days," which is the number of 24-hour periods (days) a reactor is scheduled for operation at full power output for the generation of heat energy. The number of full-power days in a reactor's operating cycle (between refueling outage times) is related to the amount of fissile uranium-235 (U-235) contained in the fuel assemblies at the beginning of the cycle. A higher percentage of U-235 in the core at the beginning of a cycle will permit the reactor to be run for a greater number of full-power days.

At the end of the operating cycle, the fuel in some of the assemblies is "spent", having spent 4 to 6 years in the reactor producing power. This spent fuel is discharged and replaced with new (fresh) fuel assemblies. Though considered "spent," these fuel assemblies contain a large quantity of fuel. In practice it is economics that determines the lifetime of nuclear fuel in a reactor. Long before all possible fission has taken place, the reactor is unable to maintain 100%, full output power, and therefore, income for the utility lowers as plant output power lowers. Most nuclear plants operate at a very low profit margin due to operating overhead, mainly regulatory costs, so operating below 100% power is not economically viable for very long. The fraction of the reactor's fuel core replaced during refueling is typically one-third, but depends on how long the plant operates between refueling. Plants typically operate on 18 month refueling cycles, or 24 month refueling cycles. This means that 1 refueling, replacing only one-third of the fuel, can keep a nuclear reactor at full power for nearly 2 years. The disposition and storage of this spent fuel is one of the most challenging aspects of the operation of a commercial nuclear power plant. This nuclear waste is highly radioactive and its toxicity presents a danger for thousands of years.[26] After being discharged from the reactor, spent nuclear fuel is transferred to the on-site spent fuel pool. The spent fuel pool is a large pool of water that provides cooling and shielding of the spent nuclear fuel. Once the energy has decayed somewhat (approximately 5 years), the fuel can be transferred from the fuel pool to dry shielded casks, that can be safely stored for thousands of years. After loading into dry shielded casks, the casks are stored on-site in a specially guarded facility in impervious concrete bunkers. On-site fuel storage facilities are designed to withstand the impact of commercial airliners, with little to no damage to the spent fuel. An average on-site fuel storage facility can hold 30 years of spent fuel in a space smaller that a football field.

Not all reactors need to be shut down for refueling; for example, pebble bed reactors, RBMK reactors, molten salt reactors, Magnox, AGR and CANDU reactors allow fuel to be shifted through the reactor while it is running. In a CANDU reactor, this also allows individual fuel elements to be situated within the reactor core that are best suited to the amount of U-235 in the fuel element.

The amount of energy extracted from nuclear fuel is called its burnup, which is expressed in terms of the heat energy produced per initial unit of fuel weight. Burn up is commonly expressed as megawatt days thermal per metric ton of initial heavy metal.