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Lewis, E.E.
Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering. Funding organisation: USDOE, Washington, DC (United States)1992
Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering. Funding organisation: USDOE, Washington, DC (United States)1992
AbstractAbstract
[en] The properties of the variational nodal method for neutron transport calculations are investigated. The method is generalized for three-dimensional multigroup criticality problems in both hexagonal-z and Cartesian geometries. The method is implemented as part of the Argonne National Laboratory Code DIF3D, and applied to a series of benchmark reactor calculations. Variational nodal methods are compared of nodal transport methods based on both interface-current and discrete ordinate approximations. Model problems are used to examine the effect of running each of the three classes of nodal transport methods on computers with massively parallel architectures
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Source
Feb 1992; 5 p; CONTRACT FG02-88ER12810; OSTI as DE93002131; NTIS; INIS; US Govt. Printing Office Dep
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Report
Literature Type
Progress Report
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INIS IssueINIS Issue
AbstractAbstract
[en] The contents of the book are presented under the following chapter headings: (1) nuclear power reactor characteristics, (2) safety assessment, (3) reactor kinetics, (4) reactivity feedback effects, (5) reactivity-induced accidents, (6) fuel element behavior, (7) coolant transients, (8) loss-of-coolant accidents, (9) accident containment, and (10) releases of radioactive materials
Original Title
Book
Primary Subject
Source
1977; 644 p; John Wiley and Sons, Inc; New York; ISBN 0-471-53335-1; 

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Book
Country of publication
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INIS VolumeINIS Volume
INIS IssueINIS Issue
Carrico, C.B.; Lewis, E.E.
Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering. Funding organisation: USDOE, Washington, DC (United States)1991
Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering. Funding organisation: USDOE, Washington, DC (United States)1991
AbstractAbstract
[en] Variational nodal transport methods are generalized for the treatment of multigroup criticality problems. The generation of variational response matrices is streamlined and automated through the use of symbolic manipulation. A new red-black partitioned matrix algorithm for the solution of the within-group equations is formulated and shown to be at once both a regular matrix splitting and a synthetic acceleration method. The methods are implemented in X- Y geometry as a module of the Argonne National Laboratory code DIF3D. For few group problems highly accurate P3 eigenvalues are obtained with computing times comparable to those of an existing interface-current nodal transport method
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1991; 5 p; International topical meeting on advances in mathematics, computation and reactor physics; Pittsburgh, PA (United States); 28 Apr - 2 May 1991; CONTRACT FG02-88ER12810; W-31109-ENG-38; OSTI as DE93003462; NTIS; INIS; US Govt. Printing Office Dep
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Report
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Conference
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INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
No abstract available
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Source
19. annual meeting of the American Nuclear Society; Chicago, Illinois, USA; 10 Jun 1973; See CONF-730611-- Published in summary form only.
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Journal Article
Literature Type
Conference
Journal
Trans. Amer. Nucl. Soc; v. 16 p. 171
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INIS VolumeINIS Volume
INIS IssueINIS Issue
Snyder, B.; Lewis, E.E.
Mathematical models and computational techniques for analysis of nuclear systems1973
Mathematical models and computational techniques for analysis of nuclear systems1973
AbstractAbstract
No abstract available
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Secondary Subject
Source
American Nuclear Society. Michigan Section; p. II.56-II.88; 1973
Record Type
Report
Literature Type
Conference
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Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Hanebutte, U.R.; Lewis, E.E.
Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering. Funding organisation: USDOE, Washington, DC (United States)1991
Northwestern Univ., Evanston, IL (United States). Dept. of Mechanical Engineering. Funding organisation: USDOE, Washington, DC (United States)1991
AbstractAbstract
[en] A discrete ordinate response matrix method is formulated for the solution of neutron transport problems on massively parallel computers. The response matrix formulation eliminates iteration on the scattering source. The nodal matrices which result from the diamond-differenced equations are utilized in a factored form which minimizes memory requirements and significantly reduces the required number of algorithm utilizes massive parallelism by assigning each spatial node to a processor. The algorithm is accelerated effectively by a synthetic method in which the low-order diffusion equations are also solved by massively parallel red/black iterations. The method has been implemented on a 16k Connection Machine-2, and S8 and S16 solutions have been obtained for fixed-source benchmark problems in X--Y geometry
Primary Subject
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Source
1991; 6 p; International topical meeting on advances in mathematics, computation and reactor physics; Pittsburgh, PA (United States); 28 Apr - 2 May 1991; CONTRACT FG02-88ER12810; OSTI as DE93003463; NTIS; INIS; US Govt. Printing Office Dep
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Report
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Conference
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Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] This books presents a balanced overview of the major methods currently available for obtaining numerical solutions in neutron and gamma ray transport. It focuses on methods particularly suited to the complex problems encountered in the analysis of reactors, fusion devices, radiation shielding, and other nuclear systems. Derivations are given for each of the methods showing how the transport equation is reduced to sets of algebraic equations suitable for solution on a digital computer. The limitations of the methods and their suitability for different classes of problems are discussed in terms of computer memory, time requirements, and accuracy
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1984; 401 p; John Wiley and Sons, Inc; New York, NY (USA)
Record Type
Book
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INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
No abstract available
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Source
Joint meeting of the American Nuclear Society and the Atomic Industrial Forum and Nuclear Energy Exhibition; San Francisco, California, USA; 11 Nov 1973; See CONF-731101-- Published in summary form only.
Record Type
Journal Article
Literature Type
Conference
Journal
Trans. Amer. Nucl. Soc; v. 17 p. 228
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Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Blomquist, R.N.; Lewis, E.E.
Northwestern Univ., Evanston, IL (USA). Technological Inst1979
Northwestern Univ., Evanston, IL (USA). Technological Inst1979
AbstractAbstract
[en] The within-group even-parity neutron transport equation is formulated with complex angular and spatial trial functions and with a complex buckling approximation. Three angular trial functions are compared; finite elements, discrete ordinates, and complex spherical harmonics. When discrete ordinates or finite element basis functions are applied, the differential equations for the real and imaginary parts of the even-parity flux are coupled by the buckling vector. The spherical harmonics equations for the real and imaginary parts of the even-parity flux, however, uncouple when one particular transverse buckling direction is chosen for a two-dimensional problem. Bilinear rectangular finite elements are applied to the spatial variable, the buckled spherical harmonics angular treatment is generalized to arbitrary order, and the resulting formulation is incorporated into a multigroup formalism using a lumped source approximation. The finite element/buckled spherical harmonics multigroup neutron transport code, FESH, has been developed from these approximations, and treats either fixed source problems or criticality eigenvalue problems. The applicability of this method to problems with several thousand spatial nodes has been demonstrated by comparison of numerical results with another spherical harmonics code and with other computational methods. Buckled two-dimensional eigenvalue calculations reveal substantial improvements over the DB2 buckling method when transverse transport effects are important
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Apr 1979; 103 p; Available from NTIS., PC A06/MF A01
Record Type
Report
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Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
No abstract available
Primary Subject
Source
Joint meeting of the American Nuclear Society and the Atomic Industrial Forum and Nuclear Energy Exhibition; San Francisco, California, USA; 11 Nov 1973; See CONF-731101-- Published in summary form only.
Record Type
Journal Article
Literature Type
Conference
Journal
Trans. Amer. Nucl. Soc; v. 17 p. 235
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
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