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Non-nuclear energy

Various types of fission reactor

Fission and radiation protection
Fusion
   

How is electricity produced in a nuclear reactor?

The principles of electricity production are the same for most types of power station – whether coal, gas or nuclear. The main difference is the source of heat to produce steam to drive the turbine and the electrical generator. Heat produced in fossil-fuel-fired power stations comes from the combustion of molecules, while in a nuclear reactor core the energy comes from the fission of atom nuclei. As a result, in nuclear fission the energy release is much bigger per gram of fuel, which means that the quantities of fuel needed are much smaller.

To give an order of magnitude, a 1 000-MWe power plant operating for one year requires about 2 million tonnes of coal (depending on the quality of the coal), but only 25 tonnes of nuclear fuel. The quantities of waste produced are commensurate with the quantities of fuel introduced in the process.

What are the basic components of a nuclear fission reactor?

The following are essential components/systems of a thermal nuclear fission reactor:

  1. The fuel: the fissile material is U-235. Most reactors are using low-enriched uranium, containing in the order of 4% of U-235. Plutonium is produced in the core during the operation by the transformation of U-238, a fertile element, into Pu-239. The fuel is made of metal or oxide pellets. In some cases plutonium is added during the fabrication process of the fuel (the fuel is then called MOX – mixed-oxide fuel)
  2. Fuel cladding: a metal shell in which the fuel pellets are contained, it protects the fuel and prevents gaseous fission products from escaping
  3. A moderator: made of light elements, it slows down the fission neutrons to thermal levels in thermal neutron reactors, without unduly absorbing them. Moderation has to be avoided in fast-neutron reactors
  4. A coolant: this transports the heat generated from the core to the steam generator for driving the turbine. The coolant can also be the moderator, but not in all types of reactor
  5. Control rods: made of neutron-absorbing material, these can be moved in or out of the core to control the chain reaction and maintain the reactor at a stable power level or decrease/increase this level
  6. A pressure vessel: containing the primary coolant, this also prevents radioactive material from escaping in case of fuel-rod failures
  7. A containment structure: (concrete) this protects the public from radiation in case of a loss-of-coolant event.

The fuel matrix (the oxide pellets), the fuel cladding, the primary coolant system (including the pressure vessel), and finally the containment constitute four successive barriers against the release of radioactivity into the environment.

What is the difference between a thermal and a fast nuclear reactor?

The majority of nuclear fission reactors operating today are thermal reactors, so named because the fast neutrons released in the fission process are slowed down to thermal energies by moderators (normal or heavy water, or graphite) before they carry on with the chain reaction process.

In fast reactors, the fission reaction can be sustained by fast neutrons without the need of a moderator. This means that the reactor core is more compact and the coolant may not be a moderator, such as liquid sodium. The main advantage with fast reactors is that they are more efficient in terms of extracting energy from the fuel within a closed fuel cycle system. U-238 is transformed into plutonium by the fast neutrons. The plutonium is then extracted and recycled as fuel. This happens also to a lower extent in thermal reactors, but it is only by using fast reactors that one might look towards long-term sustainability of nuclear energy – ‘producing more fuel than burning fuel’ – thanks to a conversion ratio greater than one, when the reactor produces more plutonium from U-238 than it consumes in the chain reaction.

What is a pressurised-water reactor (PWR)?

About 60% of the world’s commercial power reactors are pressurised-water reactors (PWRs). Initially developed by the USA and the former USSR, a PWR consists of:

  1. a compact core in a pressure vessel capable of containing ordinary water at high pressure (155 bar, i.e. 155 times the atmospheric pressure – as a matter of comparison the tyres of a car have as a maximum two times the atmospheric pressure)
  2. a primary coolant system to extract the heat from the core and transport it to steam generators
  3. a secondary coolant circuit with the steam generators and the turbine/electric generator system
  4. a containment structure containing the primary system and the auxiliaries and safety systems.

Fig. 1 A Schematic of a Pressurised Water Reactor (PWR)

Schematic of a pressurised water reactor (PWR)

What is a boiling-water reactor (BWR)?

A BWR has many similarities to a PWR but there is only one circuit with water at lower pressure so that it boils in the core at about 285°C (pressure around 80 bar). The water in the top part of the core is in the form of steam. That steam passes directly to the turbines, which are thus part of the reactor circuit.

What is a gas-cooled reactor?

This is a graphite-moderated reactor cooled by pressurised carbon dioxide. There are two types, both of which were developed mostly in the UK, but only one type is still in operation: the advanced gas-cooled reactor (AGR) – a second-generation gas-cooled reactor which uses enriched (2.5-3.5%) uranium dioxide fuel in stainless-steel cladding.

The fuel pellets are similar to those used in a PWR but larger in diameter and have a central hole. Clusters of fuel elements are joined together end-to-end in stringers which are placed in vertical holes in the graphite moderator. In AGRs, emergency shutdown in diverse situations can be accomplished by nitrogen injection, which is a strong neutron absorber.

What is a pressurised-heavy-water reactor (PHWR)?

PHWRs, which were first developed by Canada, use heavy water (deuterium or D2O) as a moderator. Because they use such an effective moderator – D2O absorbs very few neutrons – these ‘CANDU’ reactors can utilise natural uranium without enrichment.

Bundles of fuel assemblies are enclosed in separate pressure tubes, which are arranged so that they lie horizontally in a tank of heavy water. Conventional absorber (control) rods are used vertically. A secondary shutdown system with gadolinium nitrate solution is also employed. Heavy water at high pressure is heated by passing over the fuel in the pressure tubes. It is pumped to a steam generator where it boils light water (H2O) in a separate circuit. The steam drives the turbines as usual to produce electricity.

What is a water-cooled graphite-moderated reactor (RBMK)?

This type of reactor was developed by the former USSR. Its overall layout is similar to that of the AGR (see above). However, it is fuelled by enriched uranium oxide and the fuel channels are separate pressure tubes containing ordinary water under high pressure, which is allowed to boil. The resulting steam/water mixture passes to a separator and the steam feeds the turbine to produce electricity as described above.

Weaknesses in the RBMK design and a series of wrong management decisions and operator actions, including the disabling of automatic shutdown mechanisms during some tests of electricity generation at low power, led to the Chernobyl accident in April 1986. Since then, the RBMK reactors have been modified and back-fitted with some improvements to make them safer.

What is a high-temperature reactor (HTR)?

Operating reactors at much higher temperatures (about 950°C or above) would permit not only higher efficiency but could also lead to wider direct applications such as hydrogen production, which is more efficient at higher temperatures. However, such high temperatures require the resolution of many technical problems. Difficulties with maintaining water in a dense state and steam and carbon dioxide reacting with graphite led to the choice of helium as the coolant. Fuel is in the form of particles with a diameter of less than 1 mm. Each has a kernel of uranium carbide containing the uranium enriched with up to 9% U-235, surrounded by layers of carbon and silicon carbide. This gives a containment for fission products which is stable up to 2 000°C.

Fuel elements are arranged in the form of pebble bed or fuel stacks with vertical coolant passages in the graphite moderator. The cylindrical core containing the fuel elements is surrounded by a moderating graphite reflector. The control rods are located vertically. Helium gas is heated by passing it over the fuel in the core, which in more advanced designs drives a gas turbine in a direct cycle to produce electricity. The direct cycle dictates that a high integrity of fuel and reactor components is ensured.

Small test HTRs have been operating for many years in Germany, the UK and elsewhere. However, recent technological developments have made these reactors more practical than in the past, and more advanced designs are now being considered in South Africa, China, Japan and the EU.

What are the different types of past, present and future reactor?

Generation-I reactors were developed in the 1950-60s as prototypes and only a few are still running today. Most reactors operating now (including those mentioned above) are generation-II reactors, developed in series on the basis of the most successful generation-I prototypes. Generation III are ‘advanced reactors’, such as the European Pressurised Water Reactor (EPR), the AP1000 of Westinghouse, and other advanced BWRs. These focus on improving safety, economics and severe accident management scenarios. More than a dozen (generation-III) advanced reactor designs are in various stages of development. Some have evolved from the PWR, BWR and CANDU designs, while others are more radical. The best-known radical new design is the pebble-bed modular reactor or HTR, which uses helium as coolant, at very high temperatures, to drive a turbine directly.

Generation-IV designs are still on the drawing board and will not be operational on a commercial basis for at least two or three decades. Presently six different systems are being developed in the framework of the Generation IV International Forum (GIF), which brings together countries with interest in these developments. Euratom is also a member of GIF. Three of the systems are fast reactors using sodium, lead or gas as coolant. One is an advanced HTR, one is a supercritical water-cooled reactor, and the last one is a molten-salt reactor concept.

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