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Part 3. Materials Development



J. R. Weir, Jr. surfaces of graphite with pyrolytic carbon is proving difficult. We are increasing our effort in t h i s area by examination of the microstructural mechanisms of failure of the sealant and are investigating other sealants such as barren salt. The resistance of small heats of Hastelloy N modified with Ti, Nb, Zr, Si, and C to irradiation-induced embrittlement continues to be studied. We are utilizing the microstructures to evaluate the type of carbide precipitate developed in the various compositions, since there is a reasonably good correlation between the existence of MC-type carbides and good radiation resistance. Our compatibility programs involve determining the corrosion resistance of Hastelloy N in fuel salt, the coolant salt (NaBF4-NaF), and steam. The possibility of using a duplex tube of nickel on the coolant-salt side and Incoloy 800 on the steam side of a steam generator is being investigated. This combination should produce the optimum corrosion resistance for both. fluids. The major effort in our work in support of chemical processing equipment involves the development of fabrication procedures to build a complicated reductive extraction processing unit of molybdenum. Procedures have been worked out for most of the components of the system. Other possible materials that appear to be candidates for this application are tantalum and graphite. We are evaluating these materials and a few brazing alloys for their resistance to bismuth corrosion.

The areas of materials research and development discussed in Part 3 include the postoperation examination of components from the MSRE, the development of radiation-resistant impermeable graphite, investigation of radiation damage and compatibility of Hastelloy N in various environments, and materials work in support of the chemical processing equipment develop ment. One of the most important problems discovered by examining components from the MSRE is a shallow intergranular penetration into the Hastelloy N that produces intergranular cracking to a depth of several mils. This phenomenon only occurs in metal exposed to the fuel salt, and it is of concern whether it is associated with a corrosion mechanism or the penetration of certain embrittling fission products. Work is also continuing on the fission product and corrosion product deposition on graphite removed from the reactor. The graphite program involves an assessment of the resistance of commercial graphites to the dimensional and structural instabilities produced by radiation damage. Recent irradiation results indicate that graphites can be fabricated at ORNL with behavior similar to the best commercial graphites. This result is encouraging in that it increases our confidence that we understand enough about the radiation damage processes to further improve the important properties of the moderator. Maintaining low permeability (during irradiation) to gaseous fission product intrusion by sealing the