Separation of Zirconium from Zircaloy-2 for Spent Nuclear Fuel Cladding Applications using a Chloride Volatility Process

Journal Article · Wed Jun 15 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:22991865

Boras Nevarez, Rosendo ^[1]

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1. University of Nevada Las Vegas, 4505 S Maryland Pkwy, Las Vegas, Nevada 89154 (United States)

The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)

• OSTI ID: 22991865
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Journal Issue: 1 Vol. 114; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ANTIMONY 125
BOILING POINTS
CESIUM 137
CHLORIDE VOLATILITY PROCESS
CHLORINATION
CLADDING
COBALT 60
IMPURITIES
IODINATION
IRRADIATION
LOW-LEVEL RADIOACTIVE WASTES
NIOBIUM 93
REDOX PROCESS
SPENT FUELS
WASTE DISPOSAL
ZIRCALOY 2
ZIRCONIUM
ZIRCONIUM CARBONATES