STP484: Effects of Irradiation in a Thermal Reactor on the Tensile Properties of Zircaloy 2 and 4 and Borated Stainless Steel

    Baroch, CJ
    Chief, Ceramics and Metallurgy Section, and engineer, Nuclear Metallurgy Group, Babcock & Wilcox Co., Lynchburg, Va.

    Munim, AV
    Engineer, Wheelabrator Co., Bombay,

    Harbinson, EN
    Chief, Ceramics and Metallurgy Section, and engineer, Nuclear Metallurgy Group, Babcock & Wilcox Co., Lynchburg, Va.

    Pages: 17    Published: Jan 1970


    Abstract

    Babcock & Wilcox conducted a series of tests to determine the effects of irradiation and thermal aging on the tensile properties of core structural materials used in a PWR. Type 304 stainless steel (containing about 250 ppm boron) and Zircaloy 2 specimens were obtained from the first core of the Consolidated Edison Indian Point reactor, and Zircaloy 4 specimens were irradiated in the Babcock & Wilcox test reactor. The Type 304 stainless steel cladding and the Zircaloy 2 channel from the Indian Point reactor had operated at about 600 and 525 F, respectively, and had achieved peak fluences of about 3×1021 n/cm2, E > 1 MeV. The Zircaloy 4 specimens were irradiated at 130, 650, and 725 F to peak fluences of about 9×1020 n/cm2, E > 1 MeV.

    The borated stainless steel specimens were tested at 70, 600, 750, 900, 1050, 1200, 1300, and 1450 F. Some of the specimens were annealed at 1832 F for 1 h to eliminate all displacement-type defects. Irradiation increased the strength of the cladding at test temperatures below 900 F, but it had little effect on the strength at higher test temperatures. At all test temperatures the ductility of the cladding was quite low—often less than 0.5 percent. Postirradiation annealing for 1 h at 1832 F returned the strength to the unirradiated condition but had little effect on the ductility.

    Specimens obtained from both the longitudinal and transverse directions of the Zircaloy 2 channel were tested at 70, 525, and 750 F. At all test temperatures the ultimate and yield strengths increased with fluence up to the peak fluence achieved in the program. The uniform elongations were all 1 percent or less, regardless of fast fluence or specimen orientation.

    Zircaloy 4 ring specimens were tested at 900, 1100, 1300, 1450, and 1600 F. About half of the tension specimens were in the mill-annealed condition, and the remainder were cold-worked. At 1100 F and above the tensile properties of cold-worked and mill-annealed Zircaloy 4 were identical. At a test temperature of 900 F, irradiation caused some changes in the tensile properties of both materials. The extent of the effect of irradiation was dependent on the level of cold work and fluence.

    Keywords:

    irradiation, zirconium alloys, reactor cores, stainless steels, Zircaloys, radiation effects, nuclear reactors, thermal reactors, pressurized water reactors, mechanical properties, nuclear energy, helium, cladding, tensile properties, tension tests


    Paper ID: STP26620S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP26620S


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