Evaluation of the irradiation performance of Zircaloy-4 clad test rod containing annular UO/sub 2/ fuel pellets (rod 79-19) (LWBR Development Program)

Technical Report · Thu Dec 01 04:00:00 EST 1966

OSTI ID:6742246

McCauley, J E; Daniel, R C; Meieran, H B; Engel, J T

A Zircaloy-4 sheathed, rod type fuel element, containing relatively high density (95.7% TD) UO/sub 2/ annular fuel pellets, was irradiated to a depletion of 13.7 x 10/sup 20/ fissions per cm/sup 3/ of 100% dense fuel at average peak pellet temperature of 1655/sup 0/F (904/sup 0/C). The specimen failed in the reactor and developed a 1/sup 3///sub 4/ inch long crack in the cladding. The depletion limit dictated by the failure of this element applies only to the indicated test environment and to a specific test design that deliberately minimized the gap between fuel and cladding. The fuel to cladding gap was made as small as possible so that the fuel swelling characteristics could be inferred from cladding dimensional measurements. It is judged that a net plastic strain leading to sheath rupture was developed in the cladding by radioinduced swelling of the ceramic pellets concomitant with the plastic strains encountered from rod bowing. Reasonable agreement was obtained between measurements and calculations of fuel and cladding deformation as a function of irradiation lifetime. These calculations were made with an analysis method incorporated in the CYGRO-2 digital program and with a consistent set of materials' properties selected to yield overall agreement with these and other bulk oxide irradiation test results. The materials' properties and other input parameters and assumptions are presented in detail. An extensive metallographic evaluation was completed of the fuel and cladding subsequent to element failure. Theoretical models were employed to obtain estimates of UO/sub 2/ grain growth, and as-fabricated void and gas bubble migration. Calculated behavior of these parameters was consistent with observed microstructural features in the ceramic fuel pellet. (NSA 21: 11263)

• Research Organization: Bettis Atomic Power Lab., Pittsburgh, PA (USA)
• DOE Contract Number: AT(11-1)-GEN-14
• OSTI ID: 6742246
• Report Number(s): WAPD-TM-595
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500* -- Power Reactors
Breeding
ACTINIDE COMPOUNDS
ALLOYS
BREEDER REACTORS
CHALCOGENIDES
CRACKS
FUEL ELEMENTS
FUEL PELLETS
LWBR TYPE REACTORS
OXIDES
OXYGEN COMPOUNDS
PELLETS
PERFORMANCE
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
STRAINS
THERMAL REACTORS
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCALOY 4
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS