Browsing by Author "Clancy, BE"
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- ItemADL1 - an atomic data library for use in computing the behaviour of plasma devices including fusion reactors(Australian Atomic Energy Commission, 1981-05) Clancy, BE; Cook, JL; Rose, EKA data library with self-descriptive format is presented. This library provides on a fixed temperature grid reaction rate coefficients effective degree of ionisation and data for line radiation power emission for 59 ion or neutral species. Data are presented for neutral and ionised atoms of the hydrogen isotopes and for 49 'impurity' ion species ranging from helium-3 and -4 to uranium. Data origins are also discussed.
- ItemAn analysis of power transients observed in SPERT 1 reactors(Australian Atomic Energy Commission, 1976-04) Clancy, BE; Connolly, JW; Harrington, BVThe 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.
- ItemAnalysis 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, BVAn 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.
- ItemANAUSN - a one-dimensional multigroup SN transport theory module for the AUS reactor neutronics system.(Australian Atomic Energy Commission, 1982-05) Clancy, BEANAUSN is a general purpose one-dimensional discrete ordinate transport theory program which has access to AUS datapools. Fixed source reactivity and a variety of criticality search calculations can be performed. The program can be operated as a module in the AUS scheme or as a stand-alone program.
- ItemCLUTCH - an ordinary differential equation solver for initial value problems(Australian Atomic Energy Commission, 1983-08) Clancy, BEA description is given of a computer package which provides a simple interface to the LSODE (ordinary differential equation solving) subroutines developed at Lawrence Livermore Laboratory. Standardized input and output procedures help users to obtain numerical solutions to initial value problems with minimal effort. As well as providing data for a problem, the user need only provide the source for a single FORTRAN subroutine which calculates the first derivatives of the problem's dependent variables. The package can operate in batch and interactive modes and the numerical results can be displayed either as tables or graphs by using a set of simple English Language commands.
- ItemCOMFORT - an interactive FORTRAN system for the IBM360 computer(Australian Atomic Energy Commission, 1978-12) Clancy, BECOMFORT is an interactive programming system for the IBM360 computer which allows a user to edit, save, compile, execute and monitor small to medium size FORTRAN programs from a terminal. The system includes a powerful graphics package. This report describes the system and gives instruction for its use at the Australian Atomic Energy Commission Research Establishment.
- ItemThe doppler coefficient for reactors containing thorium(Australian Atomic Energy Commission, 1960-06) Keane, A; McKay, MH; Clancy, BEThe Doppler increase in the effective resonance integral has been calculated for a model absorber which approximates the resonance structure of Th232. The results indicate that for rods of pure thorium the increase depends approximately on the square root of the temperature and very good agreement is obtained with experimental values of: 1/1 - dI/dT. The calculations also show that the addition of scattering material to the fuel rods will lessen the temperature dependence of the increase in the effective resonance integral.
- ItemEffect of missing levels on the observed channels open in neutron fission(Australian Atomic Energy Commission, 1978-03) Clancy, BE; Cook, JL; Rose, EKIt is shown that the effect of missing small fission widths in the analysis of a fission width distribution is to give an apparent number of channels open greater than the actual number open. This is demonstrated both by a numerical experiment and by analytical considerations. A set of resolution probabilities is postulated such that when the apparent distribution is calculated from the true distribution, the effective number of degrees of freedom increases by a specific amount. The theory is applied to the experimental set of fission widths for neutron fission of 235U in both the J = 3 and 4 states.
- ItemIntegrals involving doppler broadened contour functions.(Australian Atomic Energy Commission, 1964-03) Clancy, BE; Keane, AA variety of integrals arising in the study of resonance absorption have been evaluated.
- ItemIntegration - analytical and numerical. Reactor physics, mathematics and computers Summer School, January 1972.(Australian Atomic Energy Commission, 1972-01) Clancy, BESome numerical methods of evaluating definite integrals are introduced.
- ItemA 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, PLThese 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).
- ItemMULGA - 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, JPA 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.
- ItemReactors, mathematics and computers summer school(Australian Atomic Energy Commision, 1975-01) Backstrom, RP; Barry, JM; Clancy, BE; Gilbert, CP; McCulloch, DBChapter 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.
- ItemThe 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, BVSince 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.
- ItemSCORCH - a zero dimensional plasma evolution and transport code for use in small and large TOKAMAK systems(Australian Atomic Energy Commission, 1984-12) Clancy, BE; Cook, JLThe zero-dimensional code SCORCH determines number density and temperature evolution in plasmas using concepts derived from the Hinton and Hazeltine transport theory. The code uses the previously reported ADL-1 data library.
- ItemStatistical distribution functions for products of variables with a gaussian distribution with zero mean(Australian Atomic Energy Commission, 1974-09) Bertram, WK; Clancy, BE; Cook, JL; Rose, EKThe statistical distribution of a product of variables which have a Gaussian distribution is investigated. These distributions are found to be given, in general, by special functions. Expansions for these functions for small values of the variable and their asymptotic behaviour are derived. The functions are tabulated for products of up to seven variates. Some simple integrals related to the functions are given.
- ItemSUPERFIT - an interactive program for function evaluation and least-squares fitting(Australian Atomic Energy Commission, 1977-03) Clancy, BEAn interactive program package is described which provides a tool for evaluating a functional form defined by the user, for plotting it over a user-specified range, for comparing the form with experimental data specified by the user, and for carrying out least-squares fits of the functional form to the experimental data points.
- ItemZAPP - a computer program for simulation of reactor power transients(Australian Atomic Energy Commission, 1983-06) Clancy, BEThis report describes a computer program which simulates power excursions in experimental fission reactors. A point reactor kinetics model is coupled with a one-dimensional heat conduction capability which allows the code to determine reactivity feedback produced as a result of density and temperature variations within the reactor. External reactivity step insertions and ramps may be included in the calculation. A simple treatment of coolant flow is available. Test cases are provided including one which simulates a run from the SPERT series of experiments.