Effect of Irradiation on the Mechanical Properties of 19-9DL Alloy

    Published: Jan 1970

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    Austenitic stainless steels, which are used extensively in water cooled nuclear reactors, are presently being considered for use in fast reactors; however, these steels are susceptible to irradiation embrittlement and irradiation induced swelling. The modified austenitic stainless steel 19-9DL alloy, on the other hand, exhibits good creep strength at high temperatures. Babcock & Wilcox conducted an exploratory program to determine the effects of irradiation on 19-9DL alloy at temperatures of 55 C (130 F), 343 C (650 F), and 413 C (775 F). The maximum fluence in this program (9.2×1020 n/cm2, E>1 MeV) was too low to prove the suitability of the alloy's potential for fast reactor applications but did show that additional evaluation of the alloy for fast reactors was warranted. The results indicate that the alloy is suitable for use in thermal reactors, since the ductility at 413 C was greater than 10 percent at a fluence approaching 1×1021 n/cm2, E>1 MeV.


    irradiation, neutron irradiation, water cooled reactors, fast reactors (nuclear), thermal reactors, nuclear fuel cladding, creep properties, mechanical properties, yield strength, ultimate strength, ductility, elongation, annealing, solid solutions, austenitic stainless steels

    Author Information:

    Lowe, AL
    Senior materials engineerchief, Babcock & Wilcox Co., Lynchburg, Va.

    Baroch, CJ
    Senior materials engineerchief, Babcock & Wilcox Co., Lynchburg, Va.

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP26646S

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