UO$sub 2$ IRRADIATIONS OF SHORT DURATION

Technical Report · Wed Apr 01 04:00:00 EST 1959

OSTI ID:4243526

Bain, A S; Robertson, J A.L.

The ''Hydraulic Rabbit'' in the NRX Reactor provides the means for subjecting fuel specimens to irradiations of short, and controlled duration. This paper describes the observations made on 47 UO/sub 2/ specimens, prepared by different methods. These experiments illustrate the value of the ''Hydraulic Rabbit'' and indicate some of the problems that can be investigated using it. Its ability to provide short irradiations makes it complementary to the conventional ''loops'' for testing fuel elements. To some extent the present specimens must be regarded as exploratory, providing necessary experience in the application of such a method. However, the marked difference in appearance of UO/ sub 2/ due to melting has been demonstrated, and the heat ratings associated with melting and grain growth have been obtained. The appearance of the irradiated oxide has been interpreted to show that a) for the dimensions and heat ratings studied, variations in the assembled diametral clearance of 0.003 to 0.014 in. have very little effect on the fuel temperature, and b) the thermal conductivity of the UO/sub 2/ does not change during irradiations of 1 to 40 minutes. (auth) 1568O Tensile specimens of annealed, 13.1% cold-worked, and tempered 25.5% cold- worked Zircaoy-2 were irradiated at 220 and 280 deg C with integrated fast- neutron fluxes of 3.6 x l0/sup 19/n/cm/sup 2/ and 2.7 x l0/sup 20/n/cm/sup 2/, respectively. Post-irradiation tensile tests performed at room temperature and 280 deg C showed that considerable irradiation hardening occurred in all the irradiated material. This was characterized by an increase in the proportional limit, yield stress, and ultimate tensile strength, and a decrease in the total and uniform per cent elongations. The amount of irradiation hardening increased with increasing fast-neutron flux, but appeared to saturate before a fast-neutron flux of 2.7 x ties as a result of the irradiation were greater in the annealed material than in the cold-worked material. The tensile properties of irradiated 13.1% cold-worked material were almost identical to those for the irradiated tempered 25.5% cold-worked material for both levels of irradiation. A yield point was developed in irradiated annealed material tested at 280 deg C whereas no yield point was present in the same material tested at room temperature. (auth)

• Research Organization: Atomic Energy of Canada Ltd., Chalk River, Ont.
• NSA Number: NSA-13-015679
• OSTI ID: 4243526
• Report Number(s): CRFD-825; AECL-806
• Country of Publication: Canada
• Language: English

Related Subjects

FUELS
GRAIN SIZE
HEATING
HYDRAULICS
MELTING
NRX
RADIATION EFFECTS
RADIATION EFFECTS ON MATERIALS
REACTORS
THERMAL CONDUCTIVITY
URANIUM DIOXIDE