Browsing by Author "McCulloch, DB"

Now showing 1 - 9 of 9
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    Buckling and integral spectrum measurements in U235 fuelled sub-critical assemblies moderated by BeO/fertile material mixtures
    (Australian Atomic Energy Commission, 1965-12) McCulloch, DB; Duerden, P; Brittliff, E
    Exponential experiments have been carried out to give the materials buckling of a number of near—homogeneous U235/aluminium alloy fuelled systems having fertile oxides intimately mixed with the BeO moderator. Relative fission rates of U235, U233, and Pu239 were also measured in the equilibrium spectrum region of each assembly. Five assemblies having 5 w/o natural uranium oxide in BeO were investigated for a range of BeO/U235 atomic ratios from 1500: 1 to 5700: 1. A similar range covering four assemblies was examined for 5 w/o thorium oxide in BeO. A comparison of the experimental results with diffusion theory calculations is included.
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    Buckling and integral spectrum measurements in U235/BeO sub-critical assemblies
    (Australian Atomic Energy Commission, 1964-07) Duerden, P; McCulloch, DB; Brittliff, E
    The materials buckling of four BeO moderated U235/aluminium alloy fuelled systems having BeO/U235 ratios of 1465, 2930, 5860, and 8790 have been measured by the exponential method. Relative fission rates of U235, U233, and Pu239 were also measured in the equilibrium spectrum region of the same assemblies. The experiments are described in detail, and the results compared with the predictions of a simple spectrum model (Westcott) pending analysis using more complex calculatlonal models.
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    A comparison of measured and calculated 238U/235U fission rate ratios for natural U02 rod-cluster fuel elements in the Zerlina Reactor
    (Australian Atomic Energy Commission, 1972-08) Durance, G; McCulloch, DB
    Experimentally measured values of the 238U/235U fission rate ratio 628 in uranium dioxide cluster fuel elements in a heavy water moderated, light water cooled reactor lattice configuration are compared with the predictions of one-group Monte Carlo cell calculations (MONTE), corrected for the actual experimental environment. A comparison between MONTE and WIMS calculations for the experiments is also given.
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    HIFAR safety analysis: frequency and offsite consequences of fault sequences initiated by within-plant failures
    (Australian Atomic Energy Commission, 1986-05) McCulloch, DB; Corran, ER; Petersen, MCE; Nicholson, FD; Innes, RW
    HIFAR fault sequences, initiated by failures of within-plant equipment and operational procedures, are analysed using probabilistic methods, and their frequencies estimated. Sequence consequences are estimated in terms of potential radiation doses to an individual at 1.6 km radius from the reactor, expressed in terms of emergency reference levels recommended by the National Health and Medical Research Council for consideration of limited evacuation. The results show that the public risk from all such sequences is extremely low.
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    A mathematician's computer study of the reactor MOATA
    (Australian Atomic Energy Commission, 1974-01) Barry, JM; Clancy, BE; Gilbert, CP; McCulloch, DB; Pollard, JP; Sanger, PL
    These notes collect together lectures on analysis of time dependent (kinetics) experiments on the reactor MOATA. The student will be introduced to scientific problem solving through the kinetics study and he will use mathematics and computers in his analysis in much the same way as a research scientist (although on a somewhat reduced scale).
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    Measurement and calculation of slow power transients in the university training reactor, Moata
    (Australian Atomic Energy Commission, 1977-09) Connolly, JW; McCulloch, DB
    Self-limiting transients of initial period down to ~ 12 s have been measured in the AAEC's 100 kW Moata (University Training Reactor (UTR)) and the results compared with the predictions of a point-kinetics heat transfer model (ZAPP code). Although natural convection significantly affects the burst shape of such slow excursions, calculation and experiment are sufficiently in agreement to give confidence that the ZAPP model, which has been well tested against SPERT I data, can be successfully applied in safety studies for this class of reactor to predict the consequences of reactivity additions up to the onset of core melting.
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    Preliminary survey of requirements for heavy water moderated reactor lattice experiments in the proposed critical facility
    (Australian Atomic Energy Commission, 1968-02) Dalton, AW; McCulloch, DB
    Calculations have been carried out to assess the fuel and control requirements for lattice measurements in heavy water moderated reactor cells, using a 'split-table' type machine with a U235/graphite driver region. The METHUSELAH code was used to calculate cell-averaged cross-section data for the heavy-water moderated region, and GYMEA for the U235/graphite driver region and the graphite reflector. CRAM calculations in four groups were then made for critical configurations in a range of driver region compositions and assembly geometries. It was established that for the type of natural uranium steam-generating heavy water reactor lattice considered, an adequately close approximation to the equilibrium neutron spectrum for the critical SGHWR could be established over a 30cm radius at the centre of the assembly, using approximately one tonne of SGHWR type fuel, 9 kg U235 and 13 tonnes of graphite. The latter two quantities would be somewhat increased if practical packing densities for the machine were taken into account. Adequate reactivity control for operation is available by use of a practicably small number (about 5) of conveniently sized plates 'black' to thermal neutrons.
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    Reactors, mathematics and computers summer school
    (Australian Atomic Energy Commision, 1975-01) Backstrom, RP; Barry, JM; Clancy, BE; Gilbert, CP; McCulloch, DB
    Chapter 1 - Physics of reactor kinetics, Chapter 2 - Mathematics of reactors, Chapter 3 - ACL - programming, Chapter 4 - Loading and saving ACL programs on IBM360 disk storage, Chapter 5 - Analogue and hybrid computers.
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    Uranium silicide fuel elements in heavy water moderated natural uranium reactors
    (Australian Atomic Energy Commission, 1972-07) McCulloch, DB; Whitfield, JH
    Some aspects of using natural uranium silicide fuel elements in heavy water moderated reactors are studied with particular attention to reactivity limited fuel burnup. Both pressurised heavy water cooled and boiling light water cooled reactor types are considered. It is shown that with suitable redesign of the fuel element and/or pressure tube geometry to take advantage of the thermal properties of U3Si, substantial improvements in fuel burnup may be expected in comparison with similar U02 fuelled reactors.