The faults associated with the AGR are very similar to those considered for the magnox reactors. The physical characteristics of the AGR are somewhat dif — terent, and different computer codes are used. The fuel is uranium dioxide in the form of hollow cylin­drical pellets about 1 cm tong and 1 cm diameter.

These are contained within stainless steel tubes about 1 m long. Thirty-six of these steel tubes or pins are mounted in three concentric rings inside a cylindrical graphite sleeve, a little longer than the pins. The complete unit, sleeve and pins is a single fuel element.

Usually, eight elements are coupled together by a tie bar running through the axis of the element and joined to an upper unit composed of a gag unit, shield plugs and a pressure-containing closure to make up a stringer. The stringer is loaded into a single channel of the reactor. There are about 300 similar channels. The uranium is enriched in the isotope U-235 to overcome the increased neutron absorption of the stainless steel of the clad compared with magnox. As a further refinement, burnable poisons contained in toroids incorporated in the graphite sleeve have been added in the latest reactors. In this case higher en­riched uranium is used, permitting longer irradiation times, thus reducing the amount of refuelling neces­sary and improving the economics of the reactor. The moderator is graphite, although of a slightly different type to that used in magnox reactors. As the fuel is irradiated, fission products are produced, some of which escape from the molecular structure in the form of a gas causing a build-up of pressure within the pin. The fault studies for the AGR are largely con­cerned with the risk that these gaseous fission pro­ducts may be released into the coolant gas and from there into the environment, and clearly this will occur if the clad melts. It will also occur if the internal gas pressure exceeds the external pressure to an ex­tent where the pin fails. Fission products may also be released if fuel is damaged during refuelling if, for instance, it is dropped. In this case the fault studies examine the measures which need to be taken in terms of the reactor power at which the refuelling may be done and the reactor tripping arrangement, to limit the consequential release of fission products to an acceptable level.

The fault studies follow a similar pattern to that described for the magnox reactors with some refine­ments. The peak limiting clad temperature is taken to be 1350°C which is about 40°C below the actual melt­ing point. When investigating depressurisation faults it is necessary to calculate the fission product gas pressure within the pin and the strength of the clad, both of which are functions of the irradiation history of the elements of the channel being considered and of the transient temperatures. From all of these studies, the limits for full power operation and the settings of the protection equipment are derived.

One further aspect of fault study work is concerned with the handling and storage of enriched fuel out­side the reactor core. The quantity of fuel required to form a critical assembly depends upon its enrich­ment, the type and quantity of moderator available and the configuration of the fuel pellets. When han­dling fuel outside the reactor, it is conceivable that damage may occur which rearranges the pellet con­figuration or that extraneous moderating material such as oil or water may be introduced. The design of the fuel storage facilities and of the fuel route is arranged to avoid such incidents, but studies are carried out to assess the potential reactivity state for any assem­bly under all credible configurations. This study is required to show a kclr of less than 0.95 and the results are assessed and agreed by a ‘criticality panel’ of experts. Finally, a certificate is drawn up for each area where enriched fuel may be held or handled and for all operations specifying the area, the allowable quantity of fuel and the nature of the operations. These certificates are called Criticality Certificates and, after agreement by the criticality panel, are approved by those Nuclear Safety Committee members from the headquarters departments and divisions.