Volume 6, Issue 7 (July 2009)

    Irradiation-Induced Grain-Boundary Solute Segregation and Its Effect on Ductile-to-Brittle Transition Temperature in Reactor Pressure Vessel Steels

    (Received 24 June 2008; accepted 2 June 2009)

    Published Online: 2009

    CODEN: JAIOAD

      Format Pages Price  
    PDF 8 $25   ADD TO CART


    Abstract

    A study on grain-boundary segregation and embrittlement in terms of the Charpy ductile-to-brittle transition temperature (DBTT) has been performed for the neutron-irradiated A533B steels with typical contents of impurities of Japanese reactor pressure vessel ones. The neutron irradiation was conducted at 563 K to a fluence of 1.3× 1024n/m2 (E>1 MeV) using material testing reactors. The neutron irradiation induced the P and Ni segregation and the reduction in C in some cases at grain-boundaries. The increase in the P segregation at high fluence (>5×10 23n/m2, E>1 MeV) was less than 0.1 in monolayer coverage for the steels with the bulk content of P not exceeding 0.02 wt%. The hardening more strongly affected the DBTT shift than the P segregation for those steels. The reduction in segregated C that enhances the grain-boundary cohesion by neutron fluence is not large enough to cause the DBTT shift.


    Author Information:

    Nishiyama, Yutaka
    Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken

    Yamaguchi, Masatake
    Center for Computational Science and e-Systems, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken

    Onizawa, Kunio
    Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken

    Iwase, Akihiko
    Dept. of Materials Science, Osaka Prefecture Univ., Sakai-shi, Osaka

    Matsuzawa, Hiroshi
    Japan Nuclear Energy Safety Organization, Tokyo, Minato-ku


    Stock #: JAI101959

    ISSN: 1546-962X

    DOI: 10.1520/JAI101959

    ASTM International is a member of CrossRef.

    Author
    Title Irradiation-Induced Grain-Boundary Solute Segregation and Its Effect on Ductile-to-Brittle Transition Temperature in Reactor Pressure Vessel Steels
    Symposium Presented at the ASTM 24th Symposium on Effects of Radiation on Nuclear Materials and the Nuclear Fuel Cycle, 2008-06-26
    Committee E10