Perform Tests and Document Results and Analysis of Oxide Layer Effects and Comparisons
- ORNL
During the initial feasibility test using actual used nuclear fuel (UNF) cladding in FY 2012, an incubation period of 30–45 minutes was observed in the initial dry chlorination. The cladding hull used in the test had been previously oxidized in a dry air oxidation pretreatment prior to removal of the fuel. The cause of this incubation period was attributed to the resistance to chlorination of an oxide layer imparted by the dry oxidation pretreatment on the cladding. Subsequently in 2013, researchers at the Korea Atomic Energy Institute (KAERI) reported on their chlorination study [R1] on ~9-gram samples of unirradiated ZirloTM cladding tubes that had been previously oxidized in air at 500oC for various time periods to impart oxide layers of varying thickness. In early 2014, discussions with Indefinite Delivery, Indefinite Quantity (IDIQ) contracted technical consultants from Westinghouse described their previous development (and patents) [R2] on methods of chemical washing to remove some or all of the hydrous oxide layer imparted on UNF cladding during irradiation in light water reactors (LWRs) . Thus, the Oak Ridge National Laboratory (ORNL) study, described herein, was planned to extend the KAERI study on the effects of anhydrous oxide layers, but on larger ~100-gram samples of unirradiated zirconium alloy cladding tubes, and to investigate the effects of various methods of chemical pretreatment prior to chlorination with 100% chlorine on the average reaction rates and Cl2 usage efficiencies.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 1160346
- Report Number(s):
- ORNL/LTR-2014/289; R&D Project: AF5805010; FCRD-SWF-2014-000252; M3FT-14OR0308062
- Country of Publication:
- United States
- Language:
- English
Similar Records
Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding
A novel protocol to recycle zirconium from zirconium alloy cladding from used nuclear fuel rods