Graphite moderator

Chernobyl. The most weU-known graphite-moderated reactor is the infamous Chemobyl-4, in Ukraine. It suffered a devastating accident in 1986 that spread radioactivity over a wide area of Europe. 

D. R. Poulter, ed.. The Design of Gas-Cooled Graphite-Moderated Reactors, Oxford University Press, London, 1963. 

Carbides of the Actinides, Uranium, and Thorium. The carbides of uranium and thorium are used as nuclear fuels and breeder materials for gas-cooled, graphite-moderated reactors (see Nuclearreactors). The actinide carbides are prepared by the reaction of metal or metal hydride powders with carbon or preferably by the reduction of the oxides uranium dioxide [1344-57-6] UO2 tduranium octaoxide [1344-59-8], U Og, or thorium... 

A variety of graphite moderated reactor concepts have evolved since the first aircooled reactors of the 1940s. Reactors with gas, water, and molten salt coolants have been constructed and a variety of fuels, and fissile/fertile fuel mixtures, have been used. The evolution and essential features of graphite moderated power producing reactors are described here, and details of their graphites cores are given. 

The neutron dose to graphite due to irradiation is commonly reported as a time integrated flux of neutrons per unit area (or fluence) referenced to a particular neutron energy. Neutron energies greater that 50 keV, 0.1 MeV, 0.18 MeV, and 1 MeV were adopted in the past and can be readily foimd in the literature. In the U.K., irradiation data are frequently reported in fluences referenced to a standard flux spectrum at a particular point in the DIDO reactor, for which the displacement rate was measured by the nickel activation [ Ni(np) t o] reaction [equivalent DIDO nickel (EDN)]. Early on, neutron irradiation doses to the graphite moderator were reported in terms of the bum-up (energy extracted) from imit mass of the adjacent nuclear fuel, i.e., MW days per adjacent tonne of fuel, or MWd/Ate. 

Graphite will creep imder neutron irradiation and stress at temperatures where thermal creep is normally negligible. The phenomenon of irradiation creep has been widely studied because of its significance to the operation of graphite moderated fission reactors. Indeed, if irradiation induced stresses in graphite moderators could not relax via radiation creep, rapid core disintegration would result. The observed creep strain has traditionally been separated into a primary reversible component ( ,) and a secondary irreversible component (Ej), both proportional to stress and to the appropriate unirradiated elastic compliance (inverse modulus) [69]. The total irradiation-induced creep strain (ej is thus ... 

Thermal oxidation of graphite moderators is significant in several contexts. In the early air-cooled reactors the moderator temperature was low and hence the thermal oxidation rate was acceptable. However, the rate increased as the graphite became damaged by neutron irradiation. Moreover, the heat produced from the exothermic reaction... 

In support of the development of graphite moderated reactors, an enormous amount of research has been conducted on the effects of neutron irradiation and radiolytic oxidation on the structure and properties of graphites. The essential mechanisms of these phenomena are understood and the years of research have translated into engineering codes and design practices for the safe design, construction and operation of gas-cooled reactors. 

Gas-cooled, graphite moderated reactors have several significant advantages over other reactor designs by virtue of their inherent passive safety characteristics. These are the result of the large thermal mass of the graphite core, the high... 

Morgan, W.C., Nuelear graphite development, operational problems, and resolution of these problems at the Hanford produetion reactors. In Proceedings of the IAEA Specialists Meeting on Graphite Moderator Lifecycle Behavior, lAEA-TECHDOC-901, IAEA, Vienna, 1996, pp. 69 77. 

Kelly, B.T., The effect of radiolytic oxidation on the graphite moderator brick strength in advanced gas-cooled reactors, Nucl. Energy, 1984, 24(3), 265 272. 

The DOE N-Reactor is one of the plutonium production reactors located on the Hanford Reservation near Richland, Washington. It is graphite moderated, pressurized water reactors that in addition to production of special nuclear materials also provided steam to turbine generators owned by the Washington Public Power Supply System for electric power production. It began op ition in 1 is put into standby status in 1988 and closed because of similarities to Chernobyl. 

A seismic event three times or more the 0.25 g safe shutdown earthquake was the only significant mechanism found to cause oxidation of the graphite moderator stack. 

N Reactor accidents are expected at lower fuel temperatures than LWR accidents. The large thermal capacity of the graphite moderator stack, the low melting point of the fuel (1,407"K) and the GSCS contribute to lower accident temperatures which retains heavy metals in the fuel. 

Other newer designs include the advanced, gas-cooled reactor (AGR), Canadian deuterium reactor (CANDUR), sodium-cooled reactor (SCR), sodium-heated reactor (SHR), and fast breeder reactor (FBR). These reactors employ either natural or enriched uranium fuels that may be modified in some way (e.g., graphite-moderated fuels). 

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