Phase II of a re-analysis examined post-accident heat Phase 2 of a re-analysis examined post-accident heat and analyzed the effect on the loss-of-coolant accident of utilizing the available cooling. The study was limited to the Phase I results of breaks of the inlet and outlet lines which were shown to be the limiting case in terms of core heatup and fuel slumping. Results of the study included: (1) the only effect post-accident core cooling supply is that portion of the buffer seal system charging water which is injected into the reactor pressure vessel through the rod drive mechanism buffer seals; (2) if all the buffer seal orifices are plugged, the available injection rate does not prevent 100% core melting; (3) if they are open, the available injection rate limits core melting to a maximum of 8%; (4) supplements to the normal containment heat sinks provide storage for the core decay heat generated over a 237-hour period, after which one of the available means of containment heat removal must be initiated. Figures comprise over half of the report.

  • Supplemental Notes:
    • This document is available for review at the Department of Commerce Library, Main Commerce Building, Washington, D.C., under reference number UNC-5191.
  • Corporate Authors:

    United Nuclear Corporation

    Research and Engineering Center, Grasslands Road
    Elmsford, NY  United States  10523

    First Atomic Ship Transport, Incorporated

    39 Broadway
    New York, NY  United States 
  • Authors:
    • Crane, A T
    • Williams Jr, F J
  • Publication Date: 1967-12-7

Media Info

  • Features: Figures; References; Tables;
  • Pagination: 39 p.

Subject/Index Terms

Filing Info

  • Accession Number: 00027022
  • Record Type: Publication
  • Source Agency: Maritime Administration
  • Report/Paper Numbers: UNC-5191
  • Files: TRIS, USDOT
  • Created Date: Feb 14 1973 12:00AM