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Molten salt research is focus of ANS local section presentation
The American Nuclear Society’s Chicago–Great Lakes Local Section hosted a presentation on February 27 on developments at the molten salt research reactor at Abilene Christian University’s Nuclear Energy Experimental Testing (NEXT) Lab.
A recording of the presentation is available on the ANS website.
T. A. Gens, R. E. Blanco
Nuclear Science and Engineering | Volume 11 | Number 3 | November 1961 | Pages 267-273
Technical Paper | doi.org/10.13182/NSE61-A26002
Articles are hosted by Taylor and Francis Online.
A Modified Zirflex process was developed in the laboratory for dissolution of 1–10 % uranium-zirconium alloy fuels clad in Zircaloy-2 to produce a nitrate solution from which uranium can be recovered by conventional solvent extraction methods. A flowsheet is presented for dissolution of 7% uranium-zirconium alloy in 5.4 M NH4F-0.33 M NH4NO3. Enough 1 M H2O2 is added continually during dissolution to yield 0.13 M H2O2 in the final solution, neglecting the amount reacting. Dissolution of a 70-mil thick sample is complete in 1 hr. The solvent extraction feed is prepared by adding aluminum nitrate and nitric acid to the dissolver solution to yield a stable solvent extraction feed solution of 0.0075 M uranium, 0.25 M zirconium, 1 M aluminum, 2 M fluoride, and 1 M nitric acid. The off-gas is approximately 98.5% NH6, 1% H2, 0.3% O2, and 0.2% N2. Conventional stainless steel such as 309SNb or Hastelloy F appear to be suitable materials of construction with corrosion rates varying from 0.1 to 3.0 mils/month.