Browsing by Author "Harrington, BV"

Now showing 1 - 12 of 12
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    An analysis of power transients observed in SPERT 1 reactors
    (Australian Atomic Energy Commission, 1976-04) Clancy, BE; Connolly, JW; Harrington, BV
    The analytical method described in Part I of this series has been applied to the calculation of SPERT I transients performed with higher initial moderator temperatures and also to those performed in a highly undermoderated core. Reasonable agreement has been obtained between calculated and experimental burst data.
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    Analysis of power transients observed in spert i reactors, Part 1 - transients in aluminium plate-type reactors initiated at ambient temperature
    (Australian Atomic Energy Commission, 1975-03) Clancy, BE; Connolly, JW; Harrington, BV
    An investigation of SPERT I reactor reactivity feedback mechanisms has been made using the modular code AUS. Feedback terms so obtained have been used in the transient analysis code ZAPP to calculate transient behaviour for step and ramp reactivity additions. A simple model of coolant boiling has been used to analyse transients for which cladding temperatures exceed the saturation temperature of water. The generally good agreement obtained with experimental data supports the case that core only temperature coefficients are much larger than those obtained by heating core and reflector.
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    Analysis of power transients observed in the SPERT II deuterium oxide moderated close packed core
    (Australian Atomic Energy Commission, 1977-09) Connolly, JW; Harrington, BV
    Power transient behaviour of the very under-moderated SPERT II core B18/68 is analysed. The experimental conditions included core pressurisation and forced coolant flow. Generally good agreement is obtained between measured and calculated data.
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    AUS model AUSED - an editing program for AUS cross section data pools
    (Australian Atomic Energy Commission, 1976-05) Harrington, BV
    AUSED is a loading and editing module for AUS cross section data pools and is part of the AUS modular code scheme for reactor systems. The fitting of subgroup parameters for the resonance theory of the module MIRANDA has been included as an option. Emphasis has been placed on flexibility and free format style input.
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    AUS98 - the 1998 version of the AUS modular neutronics code system
    (Australian Nuclear Science and Technology Organisation, 1998-07) Robinson, GS; Harrington, BV
    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multidimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results, calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. This report gives details of all system aspects of AUS and all modules except the POW3D multidimensional diffusion module.
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    Comparison of calculations made with three time-dependent neutron codes TDA, MORSE and POW
    (Australian Atomic Energy Commission, 1974-12) McGregor, BJ; Harrington, BV
    Three computer codes were compared to determine their usefulness in analysing a pulsed neutron experiment. The codes were a Monte Carlo code (MORSE) a diffusion kinetics code (POW) and a time dependent SN code (TDA). A series of test problems were devised to progressively model the experiment. All problems assumed a spherical system with an isotropic source. The first problem had its source in the first energy group for the first nanosecond. The second problem had its source distributed in time but not distributed in energy. The third problem had a source distributed in energy and time. POW and MORSE were shown to be in good agreement, with significant differences occurring only at times when the system did not correspond with the approximations made in POW. The AAEC version of the TDA code did not handle a time-dependent source. There was also a tendency for the results beyond 50 ns to be higher than those for the other two codes for the problems having a source constant over one time interval.
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    Neutronic models for the HIFAR Reactor
    (Australian Atomic Energy Commission, 1983-09) Harrington, BV
    Standard neutronic models have been developed for the AAEC's materials testing reactor HIFAR and are available as members of a partitioned data set. The models have been used to calculate reactor physics parameters related to operation and safety. Results from the calculations are presented.
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    Neutronics study of reduced enrichment fuel for the HIFAR research reactor
    (Australian Atomic Energy Commission, 1985-06) Harrington, BV; Robinson, GS
    The neutronics consequences of using lower enrichment fuels for the research reactor HIFAR have been assessed. Comparative results include neutron flux reactivity performance plutonium production and a selection of reactivity coefficients and safety-related parameters for both high and low burn-up of the fuels considered.
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    Optimisation of an epithermal beam in HIFAR for boron neutron-capture therapy
    (Australian Nuclear Science and Technology Organisation, 1987-08) Harrington, BV
    A calculational study was undertaken to investigate the feasibility of developing an epithermal beam in the 10H horizontal facility of the HIFAR research reactor suitable for boron neutron-capture therapy of deep-seated metastatic melanoma. The filters considered were Al Al/S and AlF-3. Dose intensities and therapeutic gains in a phantom were calculated to determine optimum neutron and gamma filter lengths.
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    POW3D - neutron diffusion module of the AUS system: a user's manual
    (Australian Nuclear Science and Technology Organisation, 1996-11) Harrington, BV; Pollard, JP; Barry, JM
    POW3D is a three-dimensional neutron diffusion module of the AUS modular neutronics code system. It performs eigenvalue source of feedback-free kinetics calculations. The module includes general criticality search options and extensive editing facilities including perturbation calculations. Output options include flux or reaction rate plot files. The code permits selection from one of a variety of different solution methods (MINI ICCG or SLOR) for inner iterations with region re balance to enhance convergence. A MINI accelerated Gauss-Siedel method is used for upscatter iterations with group rebalance to enhance a convergence. Chebyshev source extrapolation is applied for outer iterations.
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    The science and engineering of HIFAR safety
    (Australian Nuclear Science and Technology Organisation, 1993-12-01) Connolly, JW; Clancy, BE; Beattie, DRH; Robinson, GS; Godfrey, RM; Harrington, BV
    Since the HIFAR Safety Document was first issued major improvements have occurred in the quality of data and in the methods of calculation which are available for deterministic analysis of the behaviour of the reactor in normal or in accident conditions. Many such analyses have been carried out but the results have been reported in a wide range of internal memoranda and in external reports. In this report the most significant of the improved methods are described and the results of some of those analyses are reviewed. Principal areas covered are reactor physics of the core and reflector the dynamics of the control systems thermal hydraulic aspects important to safety margins and the emergency core cooling system. Abnormal events discussed are inadvertent reactivity insertion sequences and the loss of coolant accident. Where possible consistent sets of data are provided for use in future analyses.
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    Twenty years of experience with the Australian modular neutronics code scheme - AUS
    (The Institution of Engineers Australia, 1994-05-01) Harrington, BV
    The AUS neutronics computational code scheme developed at ANSTO is a flexible suite of modules which been extensively tested against experimental data and in blind benchmark comparisons. Early applications of AUS included calculation of SPECT reactivity coefficients for analysis of transient experiments and gave confidence in the correctness of the code. AUS has been in continuous use at ANSTO for about 20 years, is regularly used to calculate operational and safety related data for HIFAR, and has been applied to many other neutron and photon calculations.