The low volumetric power density and low operating temperatures and pres­sures of the Magnox stations led to a search in the United Kingdom for an im­proved design. The resulting advanced gas-cooled reactor (AGR) is illustrated in Figure 2.5. In common with the Magnox reactor, the AGR uses carbon dioxide as a coolant, but the coolant pressure in the AGR is 40 bars (600 psia) and the coolant outlet temperature is 650°C. To achieve these higher temperature and pressure conditions, it was necessary to make a radical change in the design of the fuel. The fuel was changed to uranium oxide, mounted in the form of pel­lets inside thin-walled stainless steel tubes, which had small transverse ribs ma­chined on the outside (Figure 2.6). These tubes (sealed at each end) were grouped in bundles of 36 (see Figure 2.6). Since the high temperatures require the use of a stainless steel can, the can material is a significant absorber of neu­trons, unlike that in the Magnox reactor, and it is necessary to enrich the ura­nium in the fuel to about 2.3% 235U (about three times the natural 235U content). The AGR design benefited from the Magnox developments, particularly the de­sign of the gas circulation system. The steam generators were mounted inside the prestressed concrete vessel, as illustrated in Figure 2.5. Since the C02 reac­tor coolant is now at a high temperature, the steam generators can be designed to provide steam under conditions similar to those found in the most efficient fossil-fuel power plant, i. e., steam at 170 bars and 560°C. This gives the AGR a considerable advantage. Its steam cycle efficiencies are around 40%, the highest of any nuclear reactor operational at present.

Referring to Table 2.3, we see that the average volumetric power density of an AGR is around three times that of the highest-rated Magnox station. The av­erage fuel rating is also higher, by a factor of approximately 4. This leads to a more compact, capital-effective design. Nevertheless, a number of technical problems in the AGR design had to be solved. One was that the carbon dioxide coolant might react with the graphite moderator under the high temperatures and radiation fields in the reactor to produce carbon monoxide by the reaction:

C02 + C -> 2CO

which would corrode the graphite and reduce its strength. It was found that precise control of the carbon monoxide and water vapor content, together with the addition of methane in small concentrations, inhibited this reaction and minimized the rate of attack on the graphite. However, too high concentrations of methane and carbon monoxide could lead to carbon formation on the fuel elements, which would impair the heat transfer by reducing the turbulence

Double Sktared Grapeite Stove

• Improved graphite to withstand longer reactor dwell

• Modified design of graphite sleeve to improve strength

Brace

• Streamlined grids and braces to reduce pressure drop

Fuel Pins

• Strong cladding material to withstand longer reactor dwell

• Coating on pins to reduce oxidation

•

Large grained UO, fuel pellets for improved fission product retention

caused by the ribs. Fortunately, there is a range of methane and carbon monox­ide concentrations (called the coolant “window") in which the satisfactory op­eration is possible without excessive corrosion or deposition.

An alte^tive design is the so-called high-temperature gas-cooled reactor (HTGR). The use of helium rather than carbon dioxide overcomes the graphite oxidation problem. Helium is inert and consequently allows higher coolant tem­peratures. The uranium fuel is in the form of coated particles. A kernel of low-en­riched uranium carbide is coated with successive layers of pyrolytically deposited carbon and impervious silicon carbide (to retain the fission products). Two dis­tinct lines of reactor development have been pursued. One line is the so-called pebble-bed reactor, developed in Germany, whose core consists simply of a stack of graphite spheres in which the coated fuel particles are embedded. A second line of development, initiated in Europe but carried forward in the United States, is the prismatic core in which vertical replaceable graphite prisms containing graphite fuel rods (in which the coated particles are embedded) and coolant pas­sages make up the core. Typically, core power densities range between 5 and 10 ^W/m3 with helium coolant outlet temperatures up to 1000°C. A number of pro­totype HTGR plants have been built to demonstrate both the pebble-bed and the prismatic-core concepts, although no commercial power plant is currently in operation.