EXPERIMENT ON CONTINUOUS RELEASE OF FISSION GAS DURING IRRADIATION. (An Interim Report)

Technical Report · Thu Feb 23 04:00:00 EST 1961

DOI:https://doi.org/10.2172/4085252· OSTI ID:4085252

Carroll, R M; Baumann, C D

A facility was constructed in which fuel materials may be irradiated to determine their ability to retain gaseous fission prcducts during irradiation. The fuel is heated by its own fission power, which is adjusted by moving the fuel into and out of the reactor neutron flux. At constant power the temperature of the fuel is regulated by air cooling. Fission gas from the fuel is entrained in an inert-gas sweep and carried outside the reactor for analysis. The neutron flux at the fuel may be determined by argon activation. Two samples of ceramic fuel, ThO/sub 2/--UO/sub 2/ and UO/sub 2/, were irradiated in the facility. The total activity of the released fission gas from both samples was dominated by shortlived isotopes. Measurements from 1000 to 2000 deg F show that iodine, xenon, and krypton were released by a temperature-dependent process above 1300 deg F. Diffusion coefficients were calculated for the UO/sub 2/ sample. There was no change in thermal conductivity of 95% density UO/sub 2/ to 2300 Mwd/ metric ton burn-up; at 1000 to 2000 deg F, K = K/sub 0/e/sup -1.//sup 21x/ >s/sup 10-//sup 3/ (T = /sup 0/K). An in-pile oxidation of UO/sub 2/ was indicated by a stored-energy release and a large increase of fission-gas release. A 3% hydrogen addition to the sweep gas partially reduced the fuel and decreased fission-gas release. (auth)

• Research Organization: Oak Ridge National Lab., Tenn.
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-15-011714
• OSTI ID: 4085252
• Report Number(s): ORNL-3050
• Country of Publication: United States
• Language: English

Related Subjects

BURNUP
CERAMICS
DENSITY
DIFFUSION
ENERGY
HALF-LIFE
HYDROGEN
IODINE
KRYPTON
METALS, CERAMICS, AND OTHER MATERIALS
MIXING
OXIDATION
RADIOISOTOPES
THERMAL CONDUCTIVITY
THORIUM OXIDES
URANIUM DIOXIDE
XENON