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Schematic diagram of the Advanced Gas-cooled Reactor. Note that the heat exchanger is contained within the steel-reinforced concrete combined pressure vessel and radiation shield.
1. Charge tubes
2. Control rods
3. Graphite moderator
4. Fuel assemblies
5. Concrete pressure vessel and radiation shielding
6. Gas circulator
7. Water
8. Water circulator
9. Heat exchanger
10. Steam

An advanced gas-cooled reactor (AGR) is a type of nuclear reactor. These are the second generation of British gas-cooled reactors, using graphite as the neutron moderator and carbon dioxide as coolant. The AGR was developed from the Magnox reactor, operating at a higher gas temperature for improved thermal efficiency, requiring stainless steel fuel cladding to withstand the higher temperature. Because the stainless steel fuel cladding has a higher neutron capture cross section than Magnox fuel cans, enriched uranium fuel is needed, with the benefit of higher "burn ups" of 18,000 MWt-days per tonne of fuel, requiring less frequent refueling. The first prototype AGR became operational in 1962[1] but the first commercial AGR did not come on line until 1976.

All AGR power stations are configured with two reactors in a single building. Each reactor has a design thermal power output of 1,500 MWt driving a 660 MWe turbine-alternator set. Because of operational restrictions, the various AGR stations produce outputs in the range 555 MWe to 625 MWe. [1]


AGR design

AGR power station at Torness

The design of the AGR was such that the final steam conditions at the boiler stop valve were identical to that of conventional coal fired power stations, thus the same design of turbo-generator plant could be used. The mean temperature of the hot coolant leaving the reactor core was designed to be 648°C. In order to obtain these high temperatures, yet ensure useful graphite core life (graphite oxidises readily in CO2 at high temperature) a re-entrant flow of coolant at the lower boiler outlet temperature of 278°C is utilised to cool the graphite, ensuring that the graphite core temperatures do not vary too much from those seen in a Magnox station. The superheater outlet temperature and pressure were designed to be 2,485 psia and 543°C.

The fuel is uranium dioxide pellets, enriched to 2.5-3.5%, in stainless steel tubes. The original design concept of the AGR was to use a beryllium based cladding. When this proved unsuitable, the enrichment level of the fuel was raised to allow for the higher neutron capture losses of stainless steel cladding. This significantly increased the cost of the power produced by an AGR. The carbon dioxide coolant circulates through the core, reaching 640°C (1,184°F)and a pressure of around 40 bar (580 psi), and then passes through boiler (steam generator) assemblies outside the core but still within the steel lined, reinforced concrete pressure vessel. Control rods penetrate the graphite moderator and a secondary system involves injecting nitrogen into the coolant to hold the reactor down. A tertiary shutdown system which operates by injecting boron balls into the reactor has been proposed 'as retrofit to satisfy the Nuclear Installations Inspectorate’s concerns about core integrity and core restraint integrity' [2].

The AGR was designed to have a high thermal efficiency (electricity generated/heat generated ratio) of about 41%, which is better than modern pressurized water reactors which have a typical thermal efficiency of 34%[3]. This is due to the higher coolant outlet temperature of about 640 °C (1,184°F) practical with gas cooling, compared to about 325 °C (617°F) for PWRs. However the reactor core has to be larger for the same power output, and the fuel burnup ratio at discharge is lower so the fuel is used less efficiently, countering the thermal efficiency advantage [2].

Like the Magnox, CANDU and RBMK reactors, and in contrast to the light water reactors, AGRs are designed to be refuelled without being shut down first. This on-load refuelling was an important part of the economic case for choosing the AGR over other reactor types, and in 1965 allowed the CEGB and the government to claim that the AGR would produce electricity cheaper than the best coal fired power stations. However fuel assembly vibration problems arose during on-load refuelling at full power, so in 1988 full power refuelling was suspended until the mid-1990s, when further trials led to a fuel rod becoming stuck in a reactor core. Only refuelling at part load or when shut down is now undertaken at AGRs. [3]

The AGR was intended to be a superior British alternative to American light water reactor designs. It was promoted as a development of the operationally (if not economically) successful Magnox design, and was chosen from a plethora of competing British alternatives - the helium cooled High Temperature Reactor (HTR), the Steam Generating Heavy Water Reactor (SGHWR) and the Fast Breeder Reactor (FBR) - as well as the American light water pressurised and boiling water reactors (PWR and BWR) and Canadian CANDU designs. The CEGB conducted a detailed economic appraisal of the competing designs and concluded that the AGR proposed for Dungeness B would generate the cheapest electricity, cheaper than any of the rival designs and the best coal fired stations.

There were great hopes for the AGR design. An ambitious construction programme of five twin reactor stations, Dungeness B, Hinckley Point B, Hunterston B, Hartlepool and Heysham was quickly rolled out, and export orders were eagerly anticipated. However, the AGR design proved to be over complex and difficult to construct on site. Notoriously bad labour relations at the time added to the problems. The lead station, Dungeness B was ordered in 1965 with a target completion date of 1970. After problems with nearly every aspect of the reactor design it finally began generating electricity in 1983, 13 years late. The follow on stations all experienced similar problems and delays. The financing cost of the capital expended, and the cost of providing replacement electricity during the delays, were enormous, totally invalidating the pre-construction economic case.

The small-scale prototype AGR at the Sellafield (Windscale) site is in the process of being decommissioned. This project is also a study of what is required to decommission a nuclear reactor safely.

Current AGR reactors

The two power stations with four AGRs at Heysham

Currently there are seven nuclear generating stations each with two operating AGRs in the United Kingdom, owned and operated by British Energy:

AGR Power Station MWe Construction started Connected to grid Commercial operation Accounting closure date
Dungeness B 1110 1965 1983 1985 2018
Hartlepool 1210 1968 1983 1989 2014
Heysham 1 1150 1970 1983 1989 2014
Heysham 2 1250 1980 1988 1989 2023
Hinkley Point B 1220 1967 1976 1976 2016
Hunterston B 1190 1967 1976 1976 2016
Torness 1250 1980 1988 1988 2023

In 2005 British Energy announced a 10-year life extension at Dungeness B, that will see the station continue operating until 2018,[4] and in 2007 announced a 5-year life extension of Hinkley Point B and Hunterston B until 2016.[5] Life extensions at other AGRs will be considered at least three years before their scheduled closure dates.

Since 2006 Hinkley Point B and Hunterston B have been restricted to about 70% of normal MWe output because of boiler-related problems requiring that they operate at reduced boiler temperatures.[5] This output restriction is likely to remain until closure.

In 2006 AGRs made the news when documents were obtained under the Freedom of Information Act 2000 by The Guardian who claimed that British Energy were unaware of the extent of the cracking of graphite bricks in the cores of their reactors. It was also claimed that British Energy did not know why the cracking had occurred and that they were unable to monitor the cores without first shutting down the reactors. British Energy later issued a statement confirming that cracking of graphite bricks is a known symptom of extensive neutron bombardment and that they were working on a solution to the monitoring problem. Also, they stated that the reactors were examined every three years as part of "statutory outages". [4]

See also


External links



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