Results of the Irradiation of R6R018 in the Advanced Test Reactor

Technical Report · Thu Apr 01 04:00:00 EDT 2010

DOI:https://doi.org/10.2172/986947· OSTI ID:986947

Robinson, Adam B; Wachs, Daniel; Medvedev, Pavel; Clark, Curtis; Chang, Gray; Lillo, Misti; Jue, Jan-Fong; Moore, Glenn; Wight, Jared

For over 30 years the Reduced Enrichment for Research and Test Reactors (RERTR) program has worked to provide the fuel technology and analytical support required to convert research and test reactors from nuclear fuels that utilize highly enriched uranium (HEU) to fuels based on low-enriched uranium (LEU) (defined as <20% U-235). This effort is driven by a desire to minimize international civilian commerce in weapons usable materials. The RERTR fuel development program has executed a wide array of fuel tests over the last decade that clearly established the viability of research reactor fuels based on uranium-molybdenum (U-Mo) alloys. Fuel testing has included a large number of dispersion type fuels capable of providing uranium densities up to approximately 8.5 g U/cc (~1.7 g U-235/cc at 20% enrichment). The dispersion fuel designs tested are very similar to existing research test reactor fuels in that the U-Mo particles simply replace the current fuel phase within the matrix. In 2003 it became evident that the first generation U-Mo-based dispersion fuel within an aluminum matrix exhibited significant fuel performance problems at high power and burn-up. These issues have been successfully addressed with a modest modification to the matrix material composition. Testing has shown that small additions of silicon (2–5 wt%) to the aluminum (Al) matrix stabilizes the fuel performance. The fuel plate R6R018 which was irradiated in the Advanced Test Reactor (ATR) as part of the RERTR-9B experiment was part of an investigation into the role of the silicon content in the matrix. This plate consisted of a U-7Mo fuel phase dispersed in an Al-3.5Si matrix clad in Al-6061. This report outlines the fabrication history, the as fabricated analysis performed prior to irradiation, the irradiation conditions, the post irradiation examination results, and an analysis of the plates behavior.

• Research Organization: Idaho National Laboratory (INL)
• Sponsoring Organization: DOE - NA
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 986947
• Report Number(s): INL/EXT-10-18074
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ALLOYS
ALUMINIUM
Dispersion
FABRICATION
FUEL PLATES
HIGHLY ENRICHED URANIUM
IRRADIATION
MATRIX MATERIALS
MODIFICATIONS
NUCLEAR FUELS
POST-IRRADIATION EXAMINATION
R6R018
RERTR
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SILICON
TEST REACTORS
TESTING
U-Mo
URANIUM
VIABILITY
WEAPONS