THE EFFECTS OF IRRADIATION ON URANIUM-PLUTONIUM-FISSIUM FUEL ALLOYS. Final Report on Metallurgy Program 6.5.5

Technical Report · Sun Jul 01 00:00:00 EDT 1962

DOI:https://doi.org/10.2172/4785371· OSTI ID:4785371

Horak, J A; Kittel, J H; Dunworth, R J

A total of 35 specimens of U-Pu-fissium alloy and 2 specimens of U-10 wt% Pu-5 wt% Mo alloy were irradiated as a part of the fue1-alloy development program for fast breeder reactors at Argonne National Laboratory. Total atom burnups ranged from 1.0 to 1.8% at maximum fuel temperatures ranging from 230 to 470 deg C. Emphasis was placed on the EBR-II Core-III reference fuel material, which is an injection-cast, U-20 wt% Pu-10 wt% fissium alloy. lt was found that this material begins to swell catastrophically at irradiation temperatures above 370 deg C. The ability of the fuel to resist swelling did not appear to vary appreciably with minor changes in Zr or fissium content. Decreasing the Pu to 10 wt%, however, significantly improved the swelling behavior of the alloy. Both pourcast and thermally cycled material and pour-cast, extruded, and thermally cycled material appeared to be more stable under irradiation than injection-cast material. Under comparable irradiation conditions, the specimens of U-20 wt% Pu- 5 wt% Mo alloy were less dimensionally stable than the U-Pu-fissium alloys investigated. (auth)

• Research Organization: Argonne National Lab., Ill.
• Sponsoring Organization: USDOE
• DOE Contract Number: W-31-109-ENG-38
• NSA Number: NSA-16-032135
• OSTI ID: 4785371
• Report Number(s): ANL-6429
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-62
• Country of Publication: United States
• Language: English

Related Subjects

METALS, CERAMICS, AND OTHER MATERIALS
BREEDING
BURNUP
CASTING
EXPANSION
EXTRUSION
FAST NEUTRONS
FISSION PRODUCTS
FUEL ELEMENTS
FUELS
HEAT TREATMENTS
IRRADIATION
MOLYBDENUM ALLOYS
PLUTONIUM
PLUTONIUM ALLOYS
QUANTITY RATIO
RADIATION EFFECTS
REACTORS
STABILITY
TEMPERATURE
TESTING
URANIUM ALLOYS
ZIRCONIUM