Volume 9, Issue 4 (April 2012)

    A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors

    (Received 18 May 2011; accepted 14 December 2011)

    Published Online: 2012

    CODEN: JAIOAD

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    Abstract

    This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK’s Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition.


    Author Information:

    Morgan, D. O.
    TCS, Serco, Gloucester, Gloucestershire

    Robinson, A. T.
    TCS, Serco, Gloucester, Gloucestershire

    Allen, D. A.
    TCS, Serco, Gloucester, Gloucestershire

    Picton, D. J.
    TCS, Serco, Gloucester, Gloucestershire

    Thornton, D. A
    TCS, Serco, Gloucester, Gloucestershire

    Shaw, S. E.
    EDF Energy, Gloucester,


    Stock #: JAI104002

    ISSN: 1546-962X

    DOI: 10.1520/JAI104002

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    Author
    Title A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors
    Symposium , 0000-00-00
    Committee E10