Graphite molten salt reactors

Molten Salt Reactor Experiment showed that salt and graphite are compatible... 

Graphite technology and component designs developed for the HTGR and Molten Salt Reactor (MSR) designs. 

The FUJI concept was proposed in connection with the philosophy of the thorium molten salt nuclear energy synergetic system (THORIMS-NES) [XXX-4 to XXX-6], explained in more detail in Section XXX-1.5. Different from the MSBR, the FUJI is a concept of a simplified molten salt reactor without continuous chemical processing and periodic core graphite replacement, aimed at attaining near-breeder characteristics in a Th-U closed fuel cycle. 

Experiments show the non wetting behavior (Fig. 3) of the fluoride salts of interest, that molten salts will not penetrate small cracks in the graphite and that the molten salt will not contact the fuel matrix (Briggs 1963 ORNL 1964 Fontana 1970). In a classical molten salt reactor where the uranium and fission products are dissolved in the fuel salt, the fuel salt is dumped to storage tanks during shutdown. 

Molten Salt Reactor Experiment [8 MW(t)] Postirradiation Graphite ... 

For high-temperature operations, materials, and fuels are key technologies. There is a century of large-scale experience in the use of fluoride molten salts. Aluminum is made by electrolysis of a mixture of bauxite and sodium aluminum fluoride salts at 1000 C in large graphite baths. Fluoride salts are compatible with graphite fuels. A smaller nuclear experience base exists with molten fluoride salts in molten salt reactors. Nickel alloys such as modified Hastelloy-N have been qualified for service to 750 C. A number of metals and carbon-carbon composites have been identified for use at much higher temperatures however, these materials have not yet been fully developed or tested for such applications. 

D. Scott, W.P. Eatherly, Graphite and xenon behavior and their influence on molten-salt reactor design, Nucl. Appl. Technol. 8 (1970) 179—189. 

If graphite could be used as a moderator in direct contact with a molten salt, it would make possible a molten-salt reactor with a breeding ratio in excess of one (see Chapter 14). INoblems that might restrict the u.scfulness of this approach arc possible reactions of graphite and the fuel salt, penetration of the pores of the graphite by the fuel, and carburization of the nickel-alloy container. 

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. 

One synthesis approach that does not rely on CNT formation from the gas phase is molten salt synthesis. The reactor consists of a vertically oriented quartz tube that contains two graphite electrodes (i.e. anode is also the crucible) and is filled with ionic salts (e.g. LiCl or LiBr). An external furnace keeps the temperature at around 600 °C, which leads to the melting of the salt. Upon applying an electric field the ions penetrate and exfoliate the graphite cathode, producing graphene-type sheets that wrap up into CNTs on the cathode surface. Subsequently, the reactor is allowed to cool down, washed with water, and nanocarbon materials are extracted with toluene [83]. This process typically yields 20-30 % MWCNTs of low purity. 

The AHTR appears to have excellent safety attributes. The combined thermal capacity of the graphite core and the molten salt coolant pool offer a large time buffer to reactor transients. The effective transfer of heat to the reactor vessel increases the effectiveness of the RVACS and DRAGS to remove decay heat, and the excellent fission product retention characteristic of molten salt provides an extra barrier to radioactive releases. The low-pressure, chemically nonreactive coolant also greatly reduces the potential for overpressurization of the reactor containment building and provides an important additional barrier for fission product release. The most important design and safety issue with the AHTR may be the performance and reliability of the thermal blanket system, which must maintain the vessel within an acceptable temperature range. 

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