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Halogenation of used aluminum matrix test reactor fuel – a bench-scale demonstration with surrogate materials

Journal Article · · Journal of Nuclear Science and Technology (Tokyo)
 [1];  [2];  [2];  [3];  [3]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Idaho, Idaho Falls, ID (United States)
  2. Univ. of Idaho, Idaho Falls, ID (United States)
  3. Idaho National Lab. (INL), Idaho Falls, ID (United States)
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In this work, experiments with surrogate materials were performed at bench scale to demonstrate a halogenation technique applicable to treatment of used aluminum matrix test reactor fuel. The technique involves dissolution and separation of aluminum from used aluminum matrix test reactor fuel in molten-halide salt systems prior to treatment and disposition of the fuel’s uranium and fission products. Demonstration of the halogenation technique was performed with neodymium metal as a non-radiological surrogate for uranium metal. Experiments involved blending forms of aluminum and neodymium metal with ammonium and lithium chloride or ammonium and lithium bromide, which upon heating decomposed into ammonia gas and the respective hydrogen chloride or bromide gas. The latter reacted with the metals to form the respective aluminum and neodymium halides. At elevated temperatures, aluminum halides gasified away from the respective neodymium halides, which fused with their respective lithium halides. Samples of fused and distillate salts were collected and analyzed, yielding extents of aluminum removal that ranged from 94.5–98.2% for chlorination runs and 91.4–97.8% for bromination runs. No neodymium was detected in the distillate fractions. Some experiments were repeated with excess reactants, and a portion of aluminum chloride distillate was processed into a consolidated waste form.

Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE); USDOE Laboratory Directed Research and Development (LDRD) Program
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
1923675
Alternate ID(s):
OSTI ID: 23185969
Report Number(s):
INL/JOU-21-61634-Rev000
Journal Information:
Journal of Nuclear Science and Technology (Tokyo), Journal Name: Journal of Nuclear Science and Technology (Tokyo) Journal Issue: 3 Vol. 59; ISSN 0022-3131
Publisher:
Taylor & FrancisCopyright Statement
Country of Publication:
United States
Language:
English

References (7)

Electrometallurgically treating metal, oxide, and al alloy spent nuclear fuel types journal July 1997
The thermal decomposition and thermodynamic properties of uranium pentabromide journal November 1973
Electrochemical separation of aluminum from uranium for research reactor spent nuclear fuel applications journal May 1999
Pyrochemical Reprocessing of Spent Fuel by Electrochemical Techniques Using Solid Aluminium Cathodes journal January 2011
Recycling of Uranium from Uranium-Aluminium alloys by Chlorination with HCl(g) journal January 2012
Chemical Equilibrium in Complex Mixtures journal May 1958
Chlorination Reactions Applied to Reprocessing of Aluminum-Uranium Spent Nuclear Fuels journal January 1997

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Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
aluminum and neodymium separation
aluminum matrix fuel
ammonium bromide
ammonium chloride
halogenation
lithium bromide
lithium chloride