A solution procedure for the neutral transport equation in plasma slab geometry is developed. Half-angle scalar fluxes, currents and averaged cross sections are introduced to provide a convenient and simple method of calculating the neutral energy distribution as an adjunct to the neutral density calculation. A forward-backward sweep numerical solution procedure, which avoids matrix inversion, is outlined.
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Georgia Institute of Technology, School of Nuclear Engineering, Atlanta, GA (United States)
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Atlanta, Georgia
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A solution procedure for the neutral transport equation in plasma slab geometry is developed. Half-angle scalar fluxes, currents and averaged cross sections are introduced to provide a convenient and simple method of calculating the neutral energy distribution as an adjunct to the neutral density calculation. A forward-backward sweep numerical solution procedure, which avoids matrix inversion, is outlined.
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