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The world’s first man-made nuclear reactor was demonstrated at an abandoned squash court under the grandstands of the University of Chicago’s Stagg Field Stadium on December 2, 1942 by Enrico Fermi and his team of scientists. The scientists were conducting research into the fission of uranium in the race to build the world’s first atomic weapon, as part of what was later to become the Manhattan Project.
Even before research into atomic weapons began, many scientists realized the potential of nuclear fission as a source of heat energy to create steam to drive turbine generators and produce electricity. After the World War II ended, work began almost immediately on constructing nuclear reactors for the purpose of generating electricity. Since that time, several different reactor types have been developed.
A popular and useful method of categorizing reactors is according to the type of coolant used. About 80% of the commercial reactors in use at the beginning of 2003 were cooled and moderated with ordinary water and are known as light water reactors (LWRs). Of these, two major types exist — pressurized water reactors (PWRs), including a Russian variant (WER), and boiling water reactors (BWRs). The majority of the remaining 20%of reactors are cooled either by heavy water or gas. The diagram below shows how the main types of commercial reactor are distributed worldwide.
Each of the main types of commercial reactor is briefly described below with data on the number of reactors current as of January 2003.
Within each basic type there are different designs resulting from different national, manufacturer or customer requirements.
Pressurized Water Reactors (PWRs)
In pressurized water reactors (PWRs) ordinary water or “light water” is used as both coolant and moderator. The coolant is kept at high pressure (about 15.5 MPa or 2,250 psi) to keep it liquid during operation, and retained within a pressure boundary comprised mainly of the reactor pressure vessel and piping in the primary system. The coolant is circulated using powerful pumps so that the heat is transferred to boil water in a separate, secondary loop in a steam generator. The resulting steam drives the electricity-producing turbine generators
A pressurized water reactor (PWR)
VVERs
VVERs are a Russian-designed reactor found mostly in Russia, but also operating in Armenia, Bulgaria, the Czech Republic, Finland, Hungary and the Slovak Republic. The name is a Russian acronym connoting a water-cooled, water-moderated energy reactor. VVERs are, in essence, Russian-designed PWRs.
First-generation VVER (type 440/230) reactors need expensive modifications because their original designs do not correspond to contemporary practices in nuclear safety. As a result, in locations such as Bulgaria and the Slovak Republic, decisions have been made to shut down some of these units.
Boiling Water Reactors (BWRs)
Boiling water reactors (BWRs) are operating in nine countries, but the majority can be found in Japan and the United States. In a BWR, ordinary water or “light water”, acts as both coolant and moderator. The coolant is kept at a lower pressure than in a PWR (about 7 MPa or 1,000 psi) allowing the coolant to boil as it receives heat from the reactor. The resultant steam is passed directly to the turbine generators to produce electricity. While the absence of a steam generator simplifies the design, as compared with PWRs, radioactivity contaminates the electricity-generating turbine.
A boiling water reactor (BWR)
Pressurized Heavy Water Reactors (PHWRs)
In 2008, there were 27 pressurized heavy water reactors (PHWRs) operating worldwide in six countries, of which 17 are in their country of origin, Canada.* The remainder of the world’s PHWRs can be found in Argentina, India, Pakistan, the Republic of Korea and Romania. Known as CANDU (CANadian Deuterium Uranium) reactors, they use heavy water (D2O, water formed with the heavier deuterium isotope of hydrogen), as both coolant and moderator.
Heavy water allows natural uranium to be used as fuel, thereby eliminating the need, and cost, to enrich the uranium. On the other hand, the production of heavy water requires a dedicated plant to separate the D2O from ordinary water, raising the concentration of D2O from much less than 1% in its natural state to 99% in a CANDU reactor. As in a PWR, the coolant is passed through a steam generator so as to boil ordinary water in a separate loop. An advantage of the CANDU design is that refuelling can take place during operation, whereas pressurized water reactors (PWRs) and boiling water reactors (BWRs) must shut down in order to refuel. This feature allows high availability but also increases the complexity of operation.
* Three reactors are currently being refurbished — Point Lepreau, Bruce 1 and Bruce 2; another two are being placed into permanent safe storage — Pickering 2 and 3.
A CANDU pressurized heavy water reactor (PHWR)
Image courtesy of www.nuclearFAQ.ca
Gas-Cooled Reactors (GCR)
Gas-cooled reactors are only in use in the United Kingdom. There are two types, the Magnox (named from the magnesium alloy used to clad the fuel elements) and the advanced gas-cooled reactor (AGR). Both use carbon dioxide as the coolant and graphite as the moderator. The Magnox uses natural uranium as fuel and the AGR, enriched uranium. Like CANDU reactors, these designs can be refuelled on-line, with the same characteristics as stated above.
RBMK
The name is a Russian acronym meaning large power boiling reactor. Ordinary water is used as the coolant and graphite as the moderator. As with a BWR, the coolant boils as it passes through the reactor and the resultant steam is passed directly to turbine generators.
As an early design, the RBMK was often built, and some are operated, without the safety characteristics and features required elsewhere. The well-known accident at Chernobyl (Ukraine) in 1986 happened to a reactor of this type.
RBMKs are the object of special safety concerns because they cannot be upgraded to correspond to contemporary safety practices at reasonable cost.
Fast Breeder Reactors
The reactor types described above are thermal reactors, most of the fission being caused by thermal neutrons. Fast reactors are designed so as to make use of fast neutrons with much higher kinetic energies. Fast reactors release more neutrons per fission than thermal reactors, and make better use of them because the relative probability of neutron capture (compared with fission) decreases at higher neutron energies. These excess neutrons can be used to convert fertile materials, for example uranium-238 (238U) and thorium-232 (232Th), into fissile materials through neutron capture. This newly created fissile material can in turn fuel the reactor. It is possible to design reactors to produce more fuel than they consume in breeder reactors. Typically, breeder reactors are fast reactors, though designs exist that could use thermal neutrons. Fast breeder reactors, by creating fuel from non-fissile isotopes and improving the efficiency of utilization through recycling, can potentially increase available world nuclear fuel resources up to 50-fold and are therefore a key element in the sustainability of nuclear energy in the very long term. Breeder reactors have been built and operated in a number of countries, though in 2002 they were operated only in France, India, Japan and the Russian Federation.
Reactor Lifetimes
Some first-generation reactors, such as the Magnox gas-cooled reactors in the United Kingdom, are still in service, though they are approaching the end of their operational lives. Many of today's reactors were built in the 1970s and 1980s and will approach lifetimes of 40 years beginning around 2015. However, studies based on operating and materials experience have revealed no major technological issues inhibiting longer operational lives for many reactors, particularly PWRs and BWRs.
Careful monitoring of plant performance, analysis of operating experience, modernization programmes and refurbishments offer good prospects for life extensions at many plants. For example, as of October 2008, three CANDU reactors were undergoing refurbishment to extend their operational life by 25 to 30 years and beyond. Other countries such as the United States and the Russian Federation are also planning to extend the lifetimes of existing reactors. In many countries decisions on plant lifetimes are made through the periodic renewal of operating licences, which involve comprehensive safety analyses using the latest methods, information and safety requirements.
Next generation reactors will be designed to operate for longer periods up to 60 years.
Resources:
- NEA, Nuclear Energy Today, 2003, p. 17, www.nea.fr/html/pub/nuclearenergytoday/nuclear-energy-today.html
- www.cityofchicago.org/Landmarks/S/SiteNuclear.html