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Showing 1 to 20 of 33 for “"neutron diffusion"”.
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Exact and Approximated Green's Functions for Neutron Diffusion Theory
Made available in DSpace on 2017-07-06T18:58:57Z (GMT). No. of bitstreams: 3 Streibich_Matthew_W_MS.pdf: 14579807 bytes, checksum: 90defe5c6cf869d03ef95f79ed0a4455 (MD5) Streibich_Matthew_W_MS_ABS.pdf: 542939 bytes, checksum: f6101bccc7d73b530bbfaf98aef882c3 (MD5) license.txt: 4813 bytes, checksum: …
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The finite element method applied to neutron diffusion problems.
Massachusetts Institute of Technology. Dept. of Nuclear Engineering. Thesis. 1973. Nucl.E.
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A Nodal Green's Function Method for Multidimensional Neutron Diffusion Calculations
Made available in DSpace on 2014-12-14T06:26:09Z (GMT). No. of bitstreams: 1 8004214.pdf: 6014344 bytes, checksum: 8ab69c6f9773f3440f6123de5b682353 (MD5) Previous issue date: 1979
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The basis and application of perturbation theory to neutron diffusion
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1959.
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A Coarse-Mesh Nodal Method for Multigroup Multidimensional Neutron Diffusion Computations
Made available in DSpace on 2017-07-06T18:55:44Z (GMT). No. of bitstreams: 3 Ougouag_Abderrafi_Mohammed-El-Amine_MS.pdf: 56136085 bytes, checksum: 179625ed39e2fdbc18ddbea719f128f8 (MD5) Ougouag_Abderrafi_Mohammed-El-Amine_MS_ABS.pdf: 634997 bytes, checksum: 89b895aa2f221ef9a70021aabe1778eb (MD5) …
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Performance Characteristics of Specified Power Reactors in Multidimensional Neutron Diffusion Problems
The multigroup neutron diffusion equations with the constraint of specified power distributions are investigated by the application of the straight-line method which can be considered as the limiting case of zero mesh space in the finite difference method. The standard partial differential form of …
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Lie group invariant finite-difference schemes for the neutron diffusion equation
… a variety of differential equations. For the neutron diffusion equation, the typical local truncation error for standard finite difference approximation is on the order of the mesh spacing squared. To improve the accuracy of the finite difference approximation of the diffusion equation, the …
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Three dimensional heterogeneous finite element method for static multi-group neutron diffusion
Because current full-core neutronic-calculations use two-group neutron diffusion and rely on homogenizing fuel assemblies, reconstructing pin powers from such a calculation is an elaborate and not very accurate process; one which becomes more difficult with increased core heterogeneity. A …
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Higher Order Nodal Methods for Solving the Steady-State Neutron Diffusion Equation
Made available in DSpace on 2017-07-06T18:55:59Z (GMT). No. of bitstreams: 3 Rajic_Hrabri_L_MS.pdf: 30908896 bytes, checksum: 49031fbb5ce3965a84550e8ed6a14c67 (MD5) Rajic_Hrabri_L_MS_ABS.pdf: 500847 bytes, checksum: dd84aaf9af22b9252ed2a0353a415f57 (MD5) license.txt: 4813 bytes, checksum: …
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Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a …
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The Experimental Determination of Thermal and Neutron Diffusion Parameters by The Use of Empirical Functions
Made available in DSpace on 2014-12-09T19:52:20Z (GMT). No. of bitstreams: 1 6711819.pdf: 2698287 bytes, checksum: 99514c7f5be31a47404ada7ebf8c1c96 (MD5) Previous issue date: 1967
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An analytic nodal method for solving the two-group, multidimensional, static and transient neutron diffusion equations
Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1979.
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A measurement of the neutron diffusion parameters of water at different temperatures by the pulsed method
The neutron diffusion parameters of water and ice were measured by the pulsed source method at two temperatures; 1.0°C. and -19°C. Neutron pulses were obtained at one millisecond intervals by modulating the beam in a Cockcroft-Walton type accelerator. The ₁H³(d,n)₂He⁴ reaction was used to generate …
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The measurement of neutron diffusion parameters in heavy and light water concentrations by the pulsed neutron technique
Neutron diffusion parameter measurements have been made at 21°C in 20, 50, and 80 percent concentrations of heavy and light water, including pure H20 and D2 0 as moderators. Pulsed neutron techniques were used with a 250-kv Cockcroft-Walton accelerator as the pulsed source. The range of bucklings …
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Derivation of Similarith solutions for the One-Dimensional, One-Velocity Neutron Diffusion Equation Using Lie Group Transformation Theory
Made available in DSpace on 2017-07-06T18:55:28Z (GMT). No. of bitstreams: 3 Myers_William_L_MS.pdf: 26862151 bytes, checksum: fb1b194ae8d1df514a990af341ae3487 (MD5) Myers_William_L_MS_ABS.pdf: 596542 bytes, checksum: 6959b7aaf4a682d0c3b55423a0f04de3 (MD5) license.txt: 4813 bytes, checksum: …
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The Partial Current Balance Method: A Local Green's Function Technique for the Numerical Solution of Multidimensional Neutron Diffusion Problems
Made available in DSpace on 2014-12-14T06:25:55Z (GMT). No. of bitstreams: 1 7606706.pdf: 4757919 bytes, checksum: 36a8981516e64055a16e86c426d56c4d (MD5) Previous issue date: 1975
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Application of Lie Groups to Discretizing Nuclear Engineering Problems
… in one dimension for the time-independent neutron diffusion equation for Cartesian, cylindrical, and spherical geometries. This method will be expanded to constructing two-term recurrence relations for an arbitrary number of spatial regions, as well as detailing starting point values for …
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Reactor Static Optimization With Integral System Equations
… with a mathematical programming method. The neutron balances of the systems under consideration are transformed into integral equations by using Green's functions. Two-group, two-dimensional Green's functions for the neutron diffusion equations have been derived.
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Nodal Methods for Problems in Fluid Mechanics and Neutron Transport (navier-Stokes Equation, Boussinesq, Lid-Driven Cavity, Natural Thermal Convection, Diffusion Synthetic Acceleration)
… developed recently for the numerical solution of neutron diffusion and neutron transport problems.
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Development of Advanced Nodal Diffusion Methods for Modern Computer Architectures
… efficient multidimensional multigroup advanced neutron diffusion nodal methods, ILLICO, have been implemented on sequential, vector, and vector-concurrent computers. Three dimensional realistic benchmark problems can be solved in vectorized mode in less than 0.73 s (33.86 Mflops) on a Cray …
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