Nuclear power reactor

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December 6, 2009, 9:45 am

This EOE article is adapted from an information paper published by the World Nuclear Association (WNA). WNA information papers are frequently updated, so for greater detail or more up to date numbers, please see the latest version on WNA website (link at end of article).


A nuclear power reactor produces and controls the release of energy from splitting the atoms of certain elements. In a nuclear power reactor, the energy released is used as heat to make steam to generate electricity. In a research reactor the main purpose is to utilize the actual neutrons produced in the core. In most naval reactors, steam drives a turbine directly for propulsion for a ship or submarine. The principal use of a nuclear reactor is to supply electrical power to the electric grid; since the nuclear reactor is essentially a zero carbon emissions technology, its role in supplying base load power to the grid is vital.
File:250px-Nuclear power plant.jpg
A nuclear power facility.

The principles for using nuclear power to produce electricity are the same for most types of reactors. The energy released from continuous fission of the atoms of the fuel is harnessed as heat in either a gas or water and is used to produce steam. The steam is used to drive the turbines which produce electricity (as in most fossil fuel plants).

There are several components common to most types of reactors:

  • Fuel: Usually pellets of uranium oxide (UO2) arranged in tubes to form fuel rods. The rods are arranged into fuel assemblies in the reactor core.
  • Moderator: This is material which slows down the neutrons released from fission so that they cause more fission. It is usually water, but may be heavy water or graphite.
  • Control rods: These are made with neutron-absorbing material such as cadmium, hafnium or boron, and are inserted or withdrawn from the core to control the rate of reaction, or to halt it. (Secondary shutdown systems involve adding other neutron absorbers, usually as a fluid, to the system.)
  • Coolant: A liquid or gas circulating through the core so as to transfer the heat from it. In light water reactors the moderator functions also as primary coolant. Except in boiling water reactorss, there is secondary coolant circuit where the steam is made. (see also later section on primary coolant characteristics.)
  • Pressure vessel or pressure tubes: Usually a robust steel vessel containing the reactor core and moderator/coolant, but it may be a series of tubes holding the fuel and conveying the coolant through the moderator.
  • Steam generator: Part of the cooling system where the heat from the reactor is used to make steam for the turbine.
  • Containment: The structure around the reactor core which is designed to protect it from outside intrusion and to protect those outside from the effects of radiation in case of any malfunction inside. It is typically a meter-thick concrete and steel structure.

There are several different types of reactors as indicated in the following table.

Nuclear power plants in commercial operation
Reactor type Main Countries Number GWe Fuel Coolant Moderator
Pressurised Water Reactor (PWR) US, France, Japan, Russia, China
265 251.6 enriched UO2 water water
Boiling Water Reactor (BWR) US, Japan, Sweden 94 86.4 enriched UO2 water water
Gas-cooled Reactor (Magnox & AGR) UK 18
10.8 natural U (metal), enriched UO2 CO2


Pressurised Heavy Water Reactor "CANDU" (PHWR) Canada 44 24.3 natural UO2 heavy water heavy water

Light Water Graphite Reactor (RBMK)

Russia 12 12.3 enriched UO2 water


Fast Neutron Reactor (FBR) Japan, France, Russia 4 1.0 PuO2 and UO2 liquid sodium none
Russia 4 0.05 enriched UO2 water graphite
TOTAL 439 384.6
GWe = capacity in thousands of megawatts.
Source: Nuclear Engineering International Handbook 2008

Most reactors need to be shut down for refuelling, so that the pressure vessel can be opened up. In this case, refuelling is at intervals of 1-2 years, when a quarter to a third of the fuel assemblies are replaced with fresh ones. The CANDU and RBMK types have pressure tubes (rather than a pressure vessel enclosing the reactor core) and can be refuelled under load by disconnecting individual pressure tubes.

If graphite or heavy water is used as moderator, it is possible to run a power reactor on natural instead of enriched uranium. Natural uranium has the same elemental composition as when it was mined (0.7% uranium-235 (235U), over 99.2% uranium-238 (238238U)), whereas enriched uranium has the proportion of the fissile isotope 235U increased by a process called enrichment, commonly to 3.5 - 5.0%. In this case, the moderator can be ordinary water, and such reactors are collectively called light water reactors. Because the light water absorbs neutrons as well as slowing them, it is less efficient as a moderator than heavy water or graphite.

Practically all fuel is ceramic uranium oxide (UO2 with a melting point of 2800°C) and most is enriched. The fuel pellets (usually about 1 cm diameter and 1.5 cm long) are typically arranged in a long zirconium alloy (zircaloy) tube to form a fuel rod, the zirconium being hard, corrosion-resistant and permeable to neutrons . Numerous rods form a fuel assembly, which is an open lattice and can be lifted into and out of the reactor core. In the most common reactors these are about 3.5 to 4 meters long.

In a new reactor with new fuel a neutron source is needed to get the reaction going. Usually this is beryllium mixed with polonium, radium or other alpha-emitter. Alpha particles from the decay cause a release of neutrons from the beryllium as it turns to carbon-12. Restarting a reactor with some used fuel may not require this, as there may be enough neutrons to achieve criticality when control rods are removed.

Burnable poisons are often used, especially in boiling water reactors (BWRs), in fuel or coolant to even out the performance of the reactor over the time from when fresh fuel is loaded to the time of refuelling. These are neutron absorbers which decay under neutron exposure, compensating for the progressive build-up of neutron absorbers in the fuel as it is burned. The best known is gadolinium, a vital ingredient of fuel in naval reactors that are very inconvenient to refuel, compelling the design of reactors capable of running more than a decade between refuellings.

In fission, most of the neutrons are released promptly, but some are delayed. These are crucial in enabling a chain reacting system (or reactor) to be controllable and to be able to be held precisely critical.

Pressurized Water Reactor (PWR)

Figure 1: Pressurized Water Reactor

Pressurized water reactors (PWRs) are the most common type of nuclear reactor, with over 230 in use for power generation and several hundred more employed for naval propulsion. The design of PWRs originated for use as a submarine power plant. PWRs use ordinary water as both coolant and moderator. The design is distinguished by having a primary cooling circuit, which flows through the core of the reactor under very high pressure, and a secondary circuit in which steam is generated to drive a turbine.

A PWR has fuel assemblies of 200-300 rods each, arranged vertically in the core, and a large reactor would have about 150-250 fuel assemblies containing 80-100 tonnes of uranium.

Water in the reactor core reaches about 325°C; hence, it must be kept under about 150 times atmospheric pressure to prevent it from boiling. Pressure is maintained by steam in a pressurizer (see diagram). In the primary cooling circuit, the water is also the moderator—if any of it were converted to steam, the fission reaction would slow down. This negative feedback effect is one of the safety features of PWRs. The secondary shutdown system involves adding boron to the primary circuit.

The secondary circuit is under less pressure and the water boils in the heat exchangers, which are thus steam generators. The steam drives the turbine to produce electricity, and is then condensed and returned to the heat exchangers in contact with the primary circuit.

Boiling Water Reactor (BWR)

Figure 2: Boiling Water Reactor

The design of boiling water reactors (BWRs) is similar to that of the PWR, except that there is only a single circuit in which the water is at lower pressure (about 75 times atmospheric pressure) so that it boils in the core at about 285°C. The reactor is designed to operate with 12-15% of the water in the top part of the core in steam form, equipping the reactor with less moderating effect and thus efficiency.

The steam passes through drier plates (steam separators) above the core and then directly to the turbines, which are thus part of the reactor circuit. Since the water around the core of a reactor is always contaminated with traces of radionuclides, the turbine must be shielded and radiological protection provided during maintenance. The cost of this tends to balance the savings due to the simpler design. Most of the radioactivity in the water is very short-lived, so the turbine hall can be entered soon after the reactor is shut down.

A BWR fuel assembly comprises 90-100 fuel rods, and there are up to 750 assemblies in a reactor core, holding up to 140 tonnes of uranium. The secondary control system involves restricting water flow through the core so that more steam in the top part reduces moderation.

Pressurized Heavy Water Reactor (PHWR or CANDU)

Figure 3: Pressurised Heavy Water Reactor

The PHWR reactor design has been developed since the 1950s in Canada as the CANDU (CANada Deuterium Uranium), and more recently also in India. It uses natural uranium oxide (0.7% 235U) as fuel, thus requiring a more efficient moderator, in this case heavy water (deuterium oxide, D2O).

The moderator is in a large tank called a calandria, penetrated by several hundred horizontal pressure tubes that form channels for the fuel, cooled by a flow of heavy water under high pressure in the primary cooling circuit, reaching 290°C. As in the PWR, the primary coolant generates steam in a secondary circuit to drive the turbines. The pressure tube is designed so that the reactor can be refuelled progressively without shutting down, by isolating individual pressure tubes from the cooling circuit.

A CANDU fuel assembly consists of a bundle of 37 half-meter-long fuel rods (ceramic fuel pellets in zircaloy tubes) along with a support structure, with 12 bundles lying end to end in a fuel channel. Control rods penetrate the calandria vertically, and a secondary shutdown system involves adding gadolinium to the moderator. The heavy water moderator circulating through the body of the calandria vessel also yields some heat (though this circuit is not shown in the diagram in Figure 3.

Advanced Gas-cooled Reactor (AGR)

Figure 4: Advanced Gas-cooled Reactor

Advanced gas-cooled reactors (AGRs) are the second generation of British gas-cooled reactors, using graphite moderators and carbon dioxide (CO2) as a coolant. AGRs are fuelled with uranium oxide pellets, enriched to 2.5-3.5%, in stainless steel tubes. The carbon dioxide circulates through the core, reaching 650°C, and then past steam generator tubes outside the core, but still inside the concrete and steel pressure vessel. Control rods penetrate the moderator and a secondary shutdown system involves injecting nitrogen into the coolant.

The AGR was developed from the Magnox reactor, also graphite-moderated and CO2-cooled, and two of these are still operating in the UK. They use natural uranium fuel in metal form.

Light Water Graphite-moderated Reactor (RBMK)

RBMK (high-power channel reactor, reaktor bolshoy moshchnosti kanalniy), is a Soviet design developed from plutonium production reactors. RBMKs employ long (7 meter) vertical pressure tubes running through a graphite moderator, and is cooled by water that is allowed to boil in the core at 290°C, much as in a BWR. Fuel is low-enriched uranium oxide made up into fuel assemblies 3.5 meters long. With moderation largely due to the fixed graphite, excess boiling simply reduces the cooling and neutron absorbtion without inhibiting the fission reaction, and problems due to positive feedback can arise, which is why they have never been built outside the Soviet Union.

Advanced reactors

Several generations of reactors are commonly distinguished. Generation I reactors were developed in 1950-60s and relatively few are still running today. They typically ran on natural uranium fuel and were graphite-moderated. Generation II reactors are typified by the present U.S. fleet and are the most in operation elsewhere. They typically use enriched uranium fuel and are mostly cooled and moderated by water. Generation III are the Advanced Reactors, the first few of which are in operation in Japan with others under construction or ready to be ordered. They are developments of the second generation designs equipped with enhanced safety features.

Generation IV nuclear reactors are still on the drawing board and will not be operational before 2020 at the earliest, probably later. They will tend to have closed fuel cycles and burn the long-lived actinides now forming part of spent fuel, so that fission products are the only high-level waste. Many will be fast neutron reactors.

More than a dozen Generation III advanced reactor designs are in various stages of development. Some are evolutionary from the PWR, BWR and CANDU designs, some are more radical departures. The former include the Advanced Boiling Water Reactor, two of which are now operating with others under construction. The best-known innovative design is the Pebble Bed Modular Reactor, using helium as coolant, at very high temperature, to drive a turbine directly.

Considering the closed fuel cycle, Generation I-III reactors recycle plutonium (and possibly uranium), while Generation IV designs are expected to have full actinide recycling capability.

Fast neutron reactors

Fast neutron reactors (only one in commercial service) do not have a moderator and utilize fast neutrons, generating power from plutonium while making more from the 238U isotope in or around the fuel. While fast neutron reactors get more than 60 times as much energy from the original uranium compared with normal reactors, they are expensive to build. Further development of them is likely in the next decade.

Floating nuclear power plants

Apart from over 200 nuclear reactors powering various kinds of ships, Rosatom in Russia has set up a subsidiary to supply floating nuclear power plants ranging in size from 70 to 600 MWe. These will be mounted in pairson a large barge, which will be permanently moored where it is needed to supply power and possibly some desalination to a shore settlement or industrial complex. The first will have two 40 MWe reactors based on those in icebreakers and will operate at Severodvinsk, in the Archangel region. Five of the next seven will be used by Gazprom for offshore oil and gas field development and for operations on the Kola and Yamal peninsulas. One is for Pevek on the Chukotka peninsula, another for Kamchatka region, both in the far east of the country. Further far east sites being considered are Yakutia and Taimyr. Electricity cost is expected to be much lower than from present alternatives.

The Russian KLT-40S is a reactor well proven in icebreakers and now proposed for wider use in desalination and, on barges, for remote area power supply. Here a 150 MWt unit produces 35 MWe (gross) as well as up to 35 MW of heat for desalination or district heating. These are designed to run 3-4 years between refuelling and it is envisaged that they will be operated in pairs to allow for outages, with on-board refuelling capability and used fuel storage. At the end of a 12-year operating cycle the whole plant is taken to a central facility for overhaul and removal of used fuel. Two units will be mounted on a 20,000 tonne barge. A larger Russian factory-built and barge-mounted reactor is the VBER-150, of 350 MW thermal, 110 MWe. The larger VBER-300 PWR is a 325 MWe unit, originally envisaged in pairs as a floating nuclear power plant, displacing 49,000 tonnes. As a cogeneration plant it is rated at 200 MWe and 1900 GJ/hr.

Lifetime of nuclear reactors.

Most of today's nuclear plants which were originally designed for 30 or 40-year operating lives. However, with major investments in systems, structures and components lives can be extended, and in several countries there are active programs to extend operating lives. In the USA most of the more than one hundred reactors are expected to be granted licence extensions from 40 to 60 years. This justifies significant capital expenditure in upgrading systems and components, including building in extra performance margins.

Some components simply wear out, corrode or degrade to a low level of efficiency. These need to be replaced. Steam generators are the most prominent and expensive of these, and many have been replaced after about 30 years where the reactor otherwise has the prospect of running for 60 years. This is essentially an economic decision. Lesser components are more straightforward to replace as they age. In Candu reactors, pressure tube replacement has been undertaken on some plants after about 30 years operation.

A second issue is that of obsolescence. For instance, older reactors have analogue instrument and control systems. Thirdly, the properties of materials may degrade with age, particularly with heat and neutron irradiation. In respect to all these aspects, investment is needed to maintain reliability and safety. Also, periodic safety reviews are undertaken on older plants in line with international safety conventions and principles to ensure that safety margins are maintained.

See also section on aging, in Safety of nuclear power reactors article.

Primary coolants

The advent of some of the designs mentioned above provides opportunity to review the various primary coolants used in nuclear reactors:

Water or heavy water must be maintained at very high pressure (1000-2200 psi, 7-15 MPa) to enable it to function above 100ºC, as in present reactors. This has a major influence on reactor engineering. However, supercritical water around 25 MPa can give 45% thermal efficiency - as at some fossil-fuel power plants today with outlet temperatures of 600ºC, and at ultra supercritical levels (30+ MPa) 50% may be attained.

Helium must be used at similar pressure (1000-2000 psi, 7-14 MPa) to maintain sufficient density for efficient operation. Again, there are engineering implications, but it can be used in the Brayton cycle to drive a turbine directly.

Carbon dioxide was used in early British reactors and their AGRs. It is denser than helium and thus likely to give better thermal conversion efficiency. There is now interest in supercritical CO2 for the Brayton cycle.

Sodium, as normally used in fast neutron reactors, melts at 98ºC and boils at 883ºC at atmospheric pressure, so despite the need to keep it dry the engineering required to contain it is relatively modest. However, normally water/steam is used in the secondary circuit to drive a turbine (Rankine cycle) at lower thermal efficiency than the Brayton cycle.

Lead or lead-bismuth are capable of higher temperature operation. They are transparent to neutrons, aiding efficiency, and do not react with water. However, they are corrosive of fuel cladding and steels, and Pb-Bi yields Po activation products. Pb-Bi melts at 125ºC and boils at 1670ºC, Pb melts at 327ºC and boils at 1737ºC. In 1998 Russia declassified a lot of research information derived from its experience with submarine reactors, and US interest in using Pb/Pb-Bi for small reactors has increased subsequently.

Molten fluoride salt boils at 1400ºC at atmospheric pressure, so allows several options for use of the heat, including using helium in a secondary Brayton cycle with thermal efficiencies of 48% at 750°C to 59% at 1000°C, or manufacture of hydrogen.

Low-pressure liquid coolants allow all their heat to be delivered at high temperatures, since the temperature drop in heat exchangers is less than with gas coolants. Also, with a good margin between operating and boiling temperatures, passive cooling for decay heat is readily achieved.

The removal of passive decay heat is a vital feature of primary cooling systems, beyond heat transfer to do work. When the fission process stops, fission product decay continues and a substantial amount of heat is added to the core. At the moment of shutdown, this is about 6% of the full power level, but it quickly drops to about 1% as the short-lived fission products decay. This heat could melt the core of a light water reactor unless it is reliably dissipated. Typically some kind of convection flow is relied upon.

See also article on cooling power plants.

Primitive reactors

The world's oldest known nuclear reactors operated at what is now Oklo in Gabon, West Africa. About 2 billion years ago, at least 17 natural nuclear reactors achieved criticality in a rich deposit of uranium ore. Each operated at about 20 kW thermal. At that time, the concentration of 235U in all natural uranium was 3.7 percent instead of 0.7 percent as at present. 235U decays much faster than 238U, whose half-life is about the same as the age of the Earth. These natural chain reactions, started spontaneously by the presence of water acting as a moderator, continued for about 2 million years before finally dying away.

During this long reaction period, about 5.4 tonnes of fission products as well as 1.5 tonnes of plutonium together with other transuranic elements were generated in the orebody. The initial radioactive products have long since decayed into stable elements but close study of the amount and location of these has shown that there was little movement of radioactive wastes during and after the nuclear reactions. Thus, plutonium and the other transuranics remained immobile.

Further Reading

See Also


Hore-Lacy, I., & Association, W. (2009). Nuclear power reactor. Retrieved from