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    New Information on Neutron Embrittlement and Embrittlement Relief of Reactor Pressure Vessel Steels


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    Significant embrittlement of several reactor pressure vessel steels as a result of exposure to neutrons has been demonstrated by studies at the U. S. Naval Research Laboratory. Neutron embrittlement of the carbon and low alloy steels investigated has been defined in terms of increases in the nil ductility transition (NDT) temperature. Increases in the NDT of pressure vessel steels as great as 545 F have been observed. The extent of embrittlement has been shown to depend upon the neutron exposure (neutrons/cm2 >1 Mev), the type of steel, and the irradiation temperature.

    Embrittlement relief through annealing heat treatment at temperatures above the pressure vessel operating temperature has been demonstrated. Significant embrittlement relief was observed even with multiple irradiation-annealing cycles; the extent of relief being primarily dependent upon the irradiation temperature and the subsequent annealing temperature.

    Experimental data are reviewed with consideration for application to operating nuclear pressure vessel conditions. Preliminary results of power reactor surveillance and of the evaluation of one pressure vessel after nuclear service are related to experimental results. The value of extending the evaluation of surveillance specimens and of pressure vessels after removal from nuclear service is reviewed with reference to current uncertainties as to possible nuclear environmental and stress effects.

    Possibilities for using the favorable aspects of certain variable experimental factors, such as differences in embrittlement sensitivity between steels, are suggested for minimizing steel embrittlement in future reactors.

    Author Information:

    Steele, L. E.
    Heads, U.S. Naval Research Laboratory, Washington, D. C.

    Hawthorne, J. R.
    Heads, U.S. Naval Research Laboratory, Washington, D. C.

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP43535S