STP611: Mechanical Properties of Fast Reactor Fuel Cladding for Transient Analysis

    Hunter, CW
    Senior scientist and senior engineer, Hanford Engineering Development Laboratory, Westinghouse Hanford Company, Richland, Wash.

    Johnson, GD
    Senior scientist and senior engineer, Hanford Engineering Development Laboratory, Westinghouse Hanford Company, Richland, Wash.

    Pages: 18    Published: Jan 1976


    Abstract

    In order to determine mechanical behavior under various simulated reactor transient events, internally pressurized specimens of fast flux-irradiated 20 percent cold-worked Type 316 stainless steel fuel pin cladding were rapidly heated until they burst. Tests were conducted at heating rates of 10F°/s and 200°F/s with pressures of 2500 to 14 300 psi (17.2 to 98.6 MPa), resulting in failure temperatures from 1000 to 2000°F (811 to 1366 K). The specimens were taken from subassemblies irradiated in the Experimental Breeder Reactor-II at temperatures from 700 to 1300°F (644 to 978 K). Peak burnup and fluence levels ranged from 28 000 to 50 000 megawatt days per metric ton metal (MWd/MTM) and 2.5 to 4.0 × 1022 neutrons (n)/cm2 (E > 0.1 MeV), respectively.

    Irradiation degraded both the failure strain and failure strength, when the transient test conditions resulted in intergranular fracture; intergranular fracture occurred above 1000 to 1200°F (811 to 922 K), depending on strain rate. Below these temperatures the fracture mode is transgranular, so that the failure strength is not reduced.

    A correlation based on the ratio of irradiated material failure strain and failure strength was developed to describe the effects of irradiation on the mechanical properties of the cladding. The strain ratio correlation indicates that the failure strain does not further decrease beyond a fluence of 2 × 1022 n/cm2. The mechanical properties of cladding which had contained fuel during irradiation were degraded more than were the properties of unfueled material.

    Keywords:

    radiation, irradiation, stainless steels, mechanical properties, pinholes, nuclear fuel claddings, ductility, strain rate, safety analysis


    Paper ID: STP38043S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP38043S


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