Volume 9, Issue 4 (April 2012)

    Reactor Pulse-Repeatability Studies at the Annular Core Research Reactor

    (Received 17 May 2011; accepted 14 December 2011)

    Published Online: 2012

    CODEN: JAIOAD

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    Abstract

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO2–BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the 9Be(γ, n)8Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times.


    Author Information:

    DePriest, K. Russell
    Principal Member Technical Staff, Applied Nuclear Technologies, Sandia National Laboratories, Albuquerque, NM

    Trinh, Tri Q.
    Member Technical Staff, Nuclear Facility Operations, Sandia National Laboratories, Albuquerque, NM

    Luker, S. Michael
    Senior Member Technical Staff, Applied Nuclear Technologies, Sandia National Laboratories, Albuquerque, NM


    Stock #: JAI103994

    ISSN: 1546-962X

    DOI: 10.1520/JAI103994

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    Author
    Title Reactor Pulse-Repeatability Studies at the Annular Core Research Reactor
    Symposium , 0000-00-00
    Committee F34