IRRADIATION OF EXTRUSION-CLAD URANIUM-2 w/o ZIRCONIUM ALLOY FOR EBR-I, MARK III
ABS>The fuel material specified for the Mark III core of EBR-I was uranium-2 wt. % zirconium alloy coextruded with Zircaloy-2 cladding. From previous work on swaged or rolled uranium-2 wt% zirconium alloy, it was anticipated that the extruded alloy would be dimensionally unstable under irradiation unless stabilized by suitable heat treatment. In order to determine an effective heat treatment, irradiation studies were made on both clad and unclad extruded uranium-2 wt.% zirconium alloy specimens at irradiation temperature estimated at 200 to 750 deg C. The irradiation specimens included material with three different heat treatinents, selected on the basis of previous studies, and material transient melted in its cladding. For unclad specimens, it was found that the irradiation temperature strongly influenced the various irradiation growth rates resulting from different heat treatments. Growth rates of the clad specimens were relatively insensitive to either irradiation temperature or prior heat treatment An exception was the transient-melted material, which shortened under irradiation. The cladding had only limited ability to restrain the swelling rates of specimens irradiated at the more elevated temperatures. Clad transient-melted material was found to be most resistant to high-temperature swelling under irradiation. The results of the present study combined with observations in earlier investigations resulted in a recommendation that the reference heat treatment for the core consist of gamma solution at 800 deg C followed by isothermal transformation at 690 deg C. (auth)
- Research Organization:
- Argonne National Lab., Lemont, Ill.
- DOE Contract Number:
- W-31-109-ENG-38
- NSA Number:
- NSA-14-005637
- OSTI ID:
- 4209647
- Report Number(s):
- ANL-5918
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-60
- Country of Publication:
- United States
- Language:
- English
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