Browsing by Author "Pollard, JP"

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    AUS diffusion module POW checkout - 1- and 2-dimensional kinetics calculations
    (Australian Atomic Energy Commission, 1977-01) Pollard, JP
    POW is the diffusion module 'workhorse' of the AUS reactor neu-tronics modular code system; its steady state calculations have been checked out against other diffusion codes (particularly CRAM and GOG). Checkout of kinetic aspects, however, is difficult as kinetic codes are not freely available. In this report POW has been checked against three benchmark calculations as well as a calculation on the 100 kW Argonaut reactor Moata.
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    AUS diffusion neutronics module - POW3D : a mathematical description
    (Australian Atomic Energy Commission, 1987-04) Barry, JM; Pollard, JP
    The mathematical and computational structure of POW3D, a general purpose zero, one, two and three-dimensional multigroup neutron diffusion code including feedback-free kinetics, is described. The code serves as a diffusion module for the AUS nuclear code system. During the production of this efficient nuclear code, a novel mathematical approach was developed to solve the large number of linear equations involved. A description is given of this technique, the method of implicit non-stationery iteration (MINI), and the way in which it is implemented in the code. The three-dimensional implementation of several other conventional approaches to the solution of the linear systems is also discussed.
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    AUS module pow - a general purpose 0,1 and 2D, multigroup neutron diffusion code including feedback-free kinetics
    (Australian Atomic Energy Commission, 1974-02) Pollard, JP
    POW is a 'workhorse' module of the AUS system and is used mainly for the following applications: (1) 2 region resonance theory data preparation from presently available fast reactor 26- or 16-group libraries, (2) 0, 1 and 2D diffusion calculations - eigenvalue or criticality search, source or feedback-free kinetics calculations, and (3) perturbation calculations. Main features of the code include the following: (1) it will run under the site AUS system or stand alone (on an IBM360/50), (2) the code uses SLOR for inner iterations with region rebalance to enhance convergence, Gauss-Seidel for upscatter iterations with group rebalance to enhance convergence, Chebyshev source extrapolation for outer iterations and a delayed neutron integrated Crank-Nicholson method for kinetics problems, (3) free format style of input is used throughout, and (4) output options include 1 and 2D flux or reaction rate plots (on a CALCOMP plotter).
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    Calculation methods for multigroup neutron cross sections used in burnup studies
    (Australian Atomic Energy Commission, 1967-10) Pollard, JP
    Calculation of burnup of material composing a nuclear reactor is important in a feasibility study of lifetime of fuel in possible reactor systems. This work discusses (i) a basis for calculating a flux spectrum in order to obtain multigroup cross sections and (ii) a method for solving the burnup equations. Under (i) particular attention is paid to the calculation of multigroup resonance cross sections and (ii) a simple and effective analytic technique is derived- In addition, this work presents, in a consistent manner, developments in the field of reactor physics which effect the calculation of multigroup data. The physical model adopted to simplify the otherwise unmanageable equations is appropriate to a large recirculating fuel reactor operating under equilibrium conditions. The work is divided into essentially three parts (i) the basis of multigroup data (Sections 2 and 3), (ii) the calculation of multigroup data (Sections 4, 5 and 6) and (iii) the solution of burnup and multigroup flux equations (Sections 7 and 8). Considering that neutron reactions are very dependent on neutron energy it is convenient to divide the energy range into three regions typified by completely different behaviour. These are (a) the fast region (Section 4) — neutrons are produced from fission, (b) the resonance region (Section 5) — neutrons slow down in energy and (c) the thermal region (Section 6) — neutrons scatter up and down in energy until lost
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    Function arising from second order approximations to the effective resonance integral
    (Australian Atomic Energy Commission, 1965-11) McKay, MH; Pollard, JP
    A function K(θ,a1,aλ,x*0), which occurs in the evaluation of second order approximations to the effective resonance integral, is investigated. Various approximations to this function are given for a range of parameters appropriate to resonance calculations and a description is given of an economic numerical method of evaluating the function accurately.
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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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    An implicit iterative scheme for solving large systems of linear equations
    (Australian Atomic Energy Commission, 1986-12) Barry, JM; Pollard, JP
    An implicit iterative scheme for the solution of large systems of linear equations arising from neutron diffusion studies is presented. The method is applied to three-dimensional reactor studies and its performance is compared with alternative iterative approaches.
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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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    MED-records: an add database of AAEC medical records since 1966
    (Australian Atomic Engery Commission, 1986-08) Barry, JM; Pollard, JP; Tucker, AD
    Since its inception in 1958 most of the staff of the AAEC Research Establishment at Lucas Heights have had annual medical examinations. Medical information accrued since 1966 has been collected as an ADD database to allow ad hoc enquiries to be made against the data. Details are given of the database schema and numerous support routines ranging from the integrity checking of input data to analysis and plotting of the summary results.
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    MULGA - a complex of codes for the determination of multigroup averaged neutron cross section data
    (Australian Atomic Energy Commission, 1963-12) Clancy, BE; Doherty, G; Keane, A; Kletzmayr, EK; Pollard, JP
    A complex of computer programmes called MULGA is described which will produce multigroup cross sections in a format suitable for input into a selection of reactor codes. Always bearing in mind that the spatial variation of flux will frustrate any determination of "exact" cross sections the maximum accuracy has been striven for within the limitations of urgency and feasibility. The programmes; together with an associated microscopic data library tape, and a specialised monitor system, have been coded for an IBM 1620 computer with 4 magnetic tapes. The basic programmes MULGA 1 and MULGA 2 have already been adapted for an ISM 7090 and the whole series will be modified for the new site computer in 1964.
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    Numerical mathematics and Fortran. Reactor physics, mathematics and computers summer school for teachers
    (Australian Atomic Energy Commission, 1972-12) Pollard, JP
    In these notes the computer is introduced via a real problem, although in miniature, and we see how a computer becomes part of a mathematician's problem solving kit. To communicate with a computer we need a language and FORTRAN, a time-proven language, is introduced.
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    Numerical mathematics and Fortran. Reactor physics, mathematics and computers summer school, January 1972
    (Australian Atomic Energy Commission, 1972-01) Pollard, JP
    In these notes the computer is introduced via a real problem, although in miniature, and we see how a computer becomes part of a mathematician's problem solving kit. To communicate with a computer we need a language and FORTRAN, a time-proven language, is introduced.
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    PEAS - a resonance absorption programme
    (Australian Atomic Energy Commission, 1964-08) Pollard, JP
    Details are given of a 7090 FORTRAN computer programme to calculate the neutron absorption in individual symmetric, single level, Doppler broadened Breit Wigner resonances for a homogeneous system of two nuclides. The flux used in determining the resonance absorption is calculated from the numerical solution of the slowing down integral equation. A listing of the programme is provided.
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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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    SKAN - a free input labelled output variable dimensioning routine for the IBM360 computer
    (Australian Atomic Energy Commission, 1978-01) Pollard, JP
    Keyword directed free format input routine, which has been used in large neutronics codes, is described. The routine also sets up variable dimension addresses for use with side entry calls. In addition, as a program debugging aid, a labelled printer dump facility is provided. The suite of routines has been used on IBM360 computers.
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    Slowing-down spectra of neutrons in heavy water and light water mixtures
    (Australian Atomic Energy Commission, 1961-10) Duncan, ME; Hines, KC; Pollard, JP
    The slowing down spectra of neutrons are obtained for heavy water, light water, and mixtures of heavy water and light water. It is assumed that fission neutrons are produced uniformly throughout an infinite moderator and the only process considered is elastic scattering, spherically symmetric in the centre of mass system. The (n, 2n) reaction with the deuterium nucleus and absorption are assumed negligible. The average transfer cross section, fast diffusion coefficient, the slowing down area, and average velocity ratio are obtained for two—group calculations using the epi — thermal spectra.
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    Spectrum calculations for neutrons slowing down by elastic collisions
    (Australian Atomic Energy Commission, 1960-11) Pollard, JP
    A method is suggested for obtaining the collision density as a function of lethargy for neutrons slowing down in an infinite homogeneous moderator containing uniformly distributed sources of fission neutrons. The only reactions considered are elastic scattering, spherically symmetric in the centre of mass system and absorption. The solution, which is exact for single element moderators when the only neutron reaction is elastic scattering, extends the well known. Greuling-Goertzel approximation and shows that it is in error by no more than 13 per cent. The method is also exact for a mixture of nuclides, provided the cross section ratios are energy independent, and a useful approximation otherwise.
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    Subroutine MLTGRD a multigrid algorithm based on multiplicative correction and implicit non-stationary iteration
    (Australian Atomic Energy Commission, 1986-11) Barry, JM; Pollard, JP
    A FORTRAN subroutine MLTGRD is provided to solve efficiently the large systems of linear equations arising from a five-point finite difference discretisation of some elliptic partial differential equations. MLTGRD is a multigrid algorithm which provides multiplicative correction to iterative solution estimates from successively reduced systems of linear equations. It uses the method of implicit non-stationary iteration for all grid levels.
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    Subroutine SOK: iterative solution of linear equations by the method of averaging functional corrections
    (Australian Atomic Energy Commission, 1968-08) Pollard, JP
    The subroutine SOK solves a set of N simultaneous linear equations by an essentially iterative method. For the method to converge at a reasonable rate (or at all) the user must choose K(< N), the order of subsidiary equations which are to be obtained from the N given equations. The matrix of coefficients of the K subsidiary equations is inverted using the direct method of Gauss-Jordan. The method is most effective on large sparse matrices that have dominant diagonal terms and for this situation it should be possible to choose K a lot less than N. The method is most advantageous compared to other iterative methods when the trial investigation of typical matrices is worthwhile. For large matrix problems , details are given of a possible compact matrix storage arrangement. For very large matrices, which even when compacted cannot fit in core, the solution procedure is feasible, provided the matrix is available from disk or tape a column at a time. The subroutine is written in FORTRAN for the IBM 360/50 computer.
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    Summer School 1978 simulation in science with mathematics and computers
    (Australian Atomic Energy Commission, 1978-12) Pollard, JP
    These notes are for a summer school which will introduce mathematically minded year-12 high school students to simulation applied to science. The course considers the application of 1. discrete Monte Carlo techniques, and 2. continuous differential techniques to a reactor problem - the transmission of neutrons through mild steel. A considerable portion of the course is devoted to electronic computing using large scientific digital computers, minicomputers, microcomputers and hybrid, analogue-digital computers. These techniques are applied to the neutron transmission problem and other simulation processes.