Advanced Gas Reactors

Advanced gas reactors (AGRs) were designed to overcome some of the inherent limitations of the Magnox design. The main problem with the Magnox design was the low power density, pressure and operating temperatures.

The first prototype AGR was built at Windscale in 1962. The commercial AGRs that were subsequently built were twin 620-660 MW plants. Seven stations were built; these entered commercial operation in the late 1970s and 1980s. The first industrial plant was at Hinkley B commissioned in 1976. These plants ran into difficulties during their construction and design phases due to problems that were both industrial and technical. In all, three different industrial groups were commissioned with different design approaches. The Dungeness B loop is shown in Figure 1.6.

The AGR uses carbon dioxide as a coolant, like the Magnox plants, but in order to achieve higher coolant pressures (~ 40 bars) and temperatures (outlet temperatures ~ 650°C), a new fuel design was required. The fuel became uranium dioxide pellets, inside stainless steel tubes.

AGR fuel had to be enriched to about 2.3% uranium-235 in order to overcome the significant neutron absorption of the stainless steel fuel cans. With this enrichment, it was

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Figure 1.6. Dungeness B advanced gas reactor. Source: http://www. british-energy. co. uk.

possible to achieve a 3-fold increase in volumetric power density with an average fuel rating of 4-fold increase compared with the best Magnox stations.

The more onerous pressure and temperature operating conditions created difficulties for the designers associated with vibration, chemistry (corrosion) and concrete insulation problems.

In the AGR, the coolant gas is circulated from the core to steam generators. These are mounted inside the pre-stressed concrete pressure vessel. These steam generators comprised 4 or 8 steam raising units. Good efficiencies are achieved as high as 40%. The steam generators provide steam at around 170 bars and 560°C, conditions that are comparable with those in an efficient fossil fuel plant.

A problem of concern for the AGR designers was attack of the graphite moderator by the carbon dioxide gas, which could oxidise the graphite and reduce its strength. This was overcome via controlled coolant chemistry with an appropriate level of water vapour content together with a small concentration of methane. This was however a delicate balance, because too much methane could result in carbon deposition on the fuel elements and consequent degradation of heat transfer.

AGRs can be refuelled on load and the fuel can remain in the core for long periods, up to 5 years. They have high fuel efficiency, up to about 40%; they have a more efficient use of fuel compared with LWRs. The AGR has a number of inherent safety features; e. g. the graphite has a large thermal capacity in the event of a primary circuit rupture.

A disadvantage of AGRs has been the limited investment of international vendors to support their technology. This, coupled with the lack of standardisation, has led to higher capital costs. It has not competed successfully outside of the UK in comparison with the PWR and BWR.

1.4.2 High Temperature Reactors

There is clearly a strong incentive to maximise the thermodynamic efficiency of nuclear power plants and one way of achieving this is to increase the temperature of the coolant. From the early days of nuclear power there has been considerable interest in helium cooled high temperature reactors (HTRs).

A 20 MW prototype, the Dragon reactor, was built and operated at Winfrith between 1964 and 1975. The plant was operated as part of an international OECD co-operative programme. Although, there were plans for a follow-on programme to Dragon, these were not pursued.

Another 13 MW prototype, the AVR, was built in 1966 at Julich in Germany based on the ‘pebble bed’ design. In this design, the fuel consists of particles of thorium or uranium dioxide fuel surrounded by carbon. These particles are a fraction of millimetre in diameter and are bonded into balls. Following AVR, a 295 MW plant was built at Schmehausen in Germany in the early 1980s and this achieved power in 1985.

In this design, the core is filled with approximately 675,000 spherical graphite fuel particles. The helium coolant is pressurised to about 40 atm and exits the core at 750°C. Heat is transferred to water and steam, circulating in stainless steel tubes within the helium. Steam passes to the steam generator at 530°C and 181 atm.

Another model, taken forward in the US was the prismatic core design. In this design, the fuel particles are formed into cylindrical rods and placed in hexagonal graphite blocks with coolant channels. An initial 40 MW prototype designed by the General Atomic Company was built at Peach Bottom in the US. This operated from 1966 to 1974. It was followed by a 330 MW prototype at Fort Saint Vrain, which came onto the grid in 1976. Here, 10,000 fuel particles are fixed in a graphite matrix with 210 fuel channels and 108 helium channels. The helium is at 48 atm, there are 1482 fuel blocks. Somewhat higher coolant outlet temperatures were achievable with this design.

The helium-cooled reactors have a number of attractions in principle. Helium is a preferred coolant to carbon dioxide in the presence of graphite since it is inert and therefore does not oxidise graphite — a problem at higher temperatures in carbon dioxide- cooled reactors.

Another attraction of the helium-cooled reactor designs discussed above was that they could be used to produce fissile material from less useful uranium fertile material. Uranium-238 is converted to plutonium-239 and uranium-233 to thorium-232. There was, therefore, the possibility of achieving very high burn-up with targets up to 100,000 MW days tonne-1.

Difficulties were encountered in the early days of these reactors and new orders were not placed following these prototypes. However, there has been a recent revival of interest in HTRs in recent years, e. g. the ESKOM project in South Africa.

At the time of writing, there is no commercial power plant of this type in operation. However, this type of reactor is one of the designs under consideration in the US Generation IV programme. These designs are discussed in detail in subsequent chapters.

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