Browsing by Author "Robinson, GS"

Now showing 1 - 19 of 19
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    3d diffusion calculations of HIFAR including the coarse control arms and their burnup
    (Australian Nuclear Science and Technology Organisation, 1991-09) Robinson, GS
    A 3D model of HIFAR which includes the coarse control arms (CCA) has been developed which is based on a 2-group relatively coarse mesh diffusion calculation. Appropriate absorption cross sections to represent the signal arm control blades were obtained by comparison with multigroup discrete ordinates cell calculations. An integral test of the CCA worth using the model showed excellent agreement with a geometrically detailed Monte Carlo calculation. Comparison with the most recent measurement of the CCA reactivity calibration showed good agreement and in particular a constant difference of about 6 per cent between calculation and measurement in change of reactivity with arm movement over the normal operating range. Extension of the model to include the burn-up of the CCA control material has provided the first calculation-based estimates of the loss of CCA effectiveness with time. Similar estimates of the worth of europium tipped control blades and their lifetime have been made. This confirmed that blades of this type have almost identical initial reactivity worth to all-cadmium blades and that their lifetime is very much longer.
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    Approximate first collision probabilities for neutrons in cylindrical and cluster lattices
    (Australian Atomic Energy Commission, 1979-05) Robinson, GS
    Methods for calculating approximate first collision probabilities for neutrons in cylindrical and cluster lattices are presented and compared with numerical solution methods. The methods differ from those of other authors in the inclusion of anisotropic boundary conditions for both geometries. The methods, which are fast enough for routine use in multigroup and resonance subgroup calculations, have been coded in FORTRAN and included in modules of the AUS scheme for reactor neutronics calculations.
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    AUS - the Australian modular scheme for reactor neutronics computations
    (Australian Atomic Energy Commission, 1975-12) Robinson, GS
    A general description is given of the AUS modular scheme for reactor neutronics calculations. The scheme currently includes modules which provide the capacity for lattice calculations, ID transport calculations, 1 and 2D diffusion calculations (with feedback—free kinetics), and burnup calculations. Details are provided of all system aspects of AUS, but individual modules are only outlined. A complete specification is given of that part of user input which controls the calculation sequence. The report also provides sufficient details of the supervisor program and of the interface data sets to enable additional modules to be incorporated in the scheme.
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    AUS burnup module char and the associated status data pool
    (Australian Atomic Energy Commission, 1975-12) Robinson, GS
    The CHAR module of the AUS reactor neutronics scheme solves the multiregion nuclide depletion equations using an analytic method. The module obtains cross section, flux and geometry data from AUS data pools, and uses the STATUS data pool which has been designed for the storage of nuclide compositions, spatial smearing factors and other miscellaneous information.
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    AUS module MIRANDA - a data preparation code based on multiregion resonance theory
    (Australian Atomic Energy Commission, 1977-07) Robinson, GS
    MIRANDA is a data preparation module of the AUS reactor neutronics scheme and is used to prepare multigroup cross-section data which are pertinent to a particular reactor system from a general purpose multigroup library of cross sections. The cross-section library has been prepared from ENDF/B and includes temperature dependent data and resonance cross sections represented by subgroup parameters. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B. flux solution, and a group condensation facility. Interaction with other AUS modules, particularly for burnup calculations, is provided.
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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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    CHAR and BURNMAC - burnup modules of the AUS neutronics code system
    (Australian Atomic Energy Commission, 1986-03) Robinson, GS
    In the AUS neutronics code system the burnup module CHAR solves the nuclide depletion equations by an analytic technique in a number of spatial zones. CHAR is usually used as one component of a lattice burnup calculation but contains features which also make it suitable for some global burnup calculations. BURNMAC is a simple accounting module based on the assumption that cross sections for a rector zone depend only on irradiation. BURNMAC is used as one component of a global calculation in which burnup is achieved by interpolation in the cross sections produced from a previous lattice calculation.
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    A comparison of neutron resonance absorption in thermal reactor lattices in the AUS neutronics code system with Monte Carlo calculations
    (Australian Atomic Energy Commission, 1985-08) Robinson, GS
    The calculation of resonance shielding by the subgroup method as incorporated in the MIRANDA module of the AUS neutronics code system is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered AUS results for resonance absorption were high by up to two per cent.
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    EDITAR: a module for reaction rate editing and cross-section averaging within the AUS neutronics code system
    (Australian Atomic Energy Commission, 1986-03) Robinson, GS
    The EDITAR module of the AUS neutronics code system edits one and two-dimensional flux data pools produced by other AUS modules to form reaction rates for materials and their constituent nuclides and to average cross sections over space and energy. The module includes a Bsub(L) flux calculation for application to cell leakage. The STATUS data pool of the AUS system is used to enable the 'unsmearing' of fluxes and nuclide editing with minimal user input. The module distinguishes between neutron and photon groups and printed reaction rates are formed accordingly. Bilinear weighting may be used to obtain material reactivity worths and to average cross sections. Bilinear weighting is at present restricted to diffusion theory leakage estimates made using mesh-average fluxes.
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    Extension of the AUS reactor neutronics system for application to fusion blanket neutronics.
    (Australian Atomic Energy Commission, 1984-03) Robinson, GS
    The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies.
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    Generation and validation of a cross section library based on ENDF/B-VI for the AUS neutronics code system.
    (Australian Nuclear Science and Technology Organisation, 1993-12) Robinson, GS
    The generation of a cross section library with 200 neutron and 37 photon groups from ENDF/B-VI for use in the AUS modular neutronics code system is described. The NJOY code was used for most of the library preparation but a revision of previous AUS methods was used for the neutron resonance treatment. The library should be suitable for thermal and fast fission reactors fusion blankets and various neutron applications. The validity of AUS with the library was established for thermal and fast reactor systems by an extensive set of comparisons with benchmark experiments which were mainly taken from the ENDF compilation. The performance of AUS with the library was much improved over that with the previous ENDF/B-IV based library.
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    A guide to the AUS modular neutronics code system.
    (Australian Atomic Energy Commission, 1987-04) Robinson, GS
    A general description is given of the AUS modular neutronics code system which may be used for calculations of a very wide range of fission reactors fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations one-dimensional transport calculations and one two and three-dimensional diffusion calculations burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system.
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    GYMEA - a nuclide depletion, space independent, multigroup neutron diffusion, data preparation code
    (Australian Nuclear Science and Technology Organisation, 1966-03) Pollard, JP; Robinson, GS
    GYMEA is a multi-purpose neutronics code used extensively in the H.T.G.C.R. project of the A.A.E.C. The code has three main functions: (i) running of burnup surveys of near—homogeneous reactor systems, (ii) running of neutron diffusion calculations in nuclear data library correlations with measurements made in near—homogeneous subcritical assemblies carried out at the A.A.E.C, and (Hi) preparation of suitably averaged cross section data from the code's 120—group (with resonance parameters), 70—nuclide main data library for direct input to space-dependent codes such as CRAM, 4-ZOOM (9-ZOOM translated to the IBM 7040), DSN and TDC (Carlsons FLOCO versions translated to the IBM 7040), and WDSN. Because free input is used and a toy compiler is provided, new applications of the code can be developed by code users to extend its function. The source language is FORTRAN IV (95 per cent.) and MAP (5 per cent.) for the IBM 7040 (and later an IBM 360). 32K words of core storage are required together with 4—6 magnetic tapes and an on —line 1401 including reader, punch, and printer. Typical runs for data preparation take 3 minutes, diffusion calculations 10 minutes and burnup calculations 15 minutes. A few illustrative examples are included in the report.
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    ICPP - a collision probability module for the AUS neutronics code system.
    (Australian Atomic Energy Commission, 1985-10) Robinson, GS
    The isotropic collision probability program (ICPP) is a module of the AUS neutronics code system which calculates first flight collision probabilities for neutrons in one-dimensional geometries and in clusters of rods. Neutron sources including scattering are assumed to be isotropic and to be spatially flat within each mesh interval. The module solves the multigroup collision probability equations for eigenvalue or fixed source problems.
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    MCRP - a Monte Carlo resonance program for neutrons slowing down in single rod and rod cluster lattices.
    (Australian Atomic Energy commission, 1985-09) Doherty, G; Robinson, GS
    MCRP is a Monte Carlo computer program for tracking neutrons slowing down in single rod and rod cluster lattices. The code is intended for calculations of resonance absorption in reactor fuel nuclides using cross sections at 124 000 energy points below 20 keV. The only intrinsic assumptions are that scattering is both elastic and isotropic in the centre of mass system.
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    MIRANDA - module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system.
    (Australian Atomic Energy Commission, 1985-12) Robinson, GS
    MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab cylinder or cluster geometry a homogeneous B sub L flux solution and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual.
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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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    A reactor physics survey of water moderated, uranium-aluminium plate fuelled research reactors for a range of uranium enrichment
    (Australian Atomic Energy Commission, 1979-07) Robinson, GS
    Results obtained from reactor physics calculations of water moderated research reactors with fuel in the form of uranium-aluminium alloy plates are presented. The major parameters considered are uranium enrichment, uranium weight per cent in the fuel meat, 35U loading per plate and the water gap between plates. The calculations are based on the SILOE reactor and particular emphasis is placed on the requirements for a proposed new AAEC research reactor. Sufficient detail is given to enable the determination of core size, fuel consumption rate and neutron flux levels in the core and reflector when this study is combined with a thermal-hydraulic analysis.
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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.