Abstract
Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.
Citation Formats
Carlvik, I.
Calculations of Neutron Flux Distributions by Means of Integral Transport Methods.
Sweden: N. p.,
1967.
Web.
Carlvik, I.
Calculations of Neutron Flux Distributions by Means of Integral Transport Methods.
Sweden.
Carlvik, I.
1967.
"Calculations of Neutron Flux Distributions by Means of Integral Transport Methods."
Sweden.
@misc{etde_20956318,
title = {Calculations of Neutron Flux Distributions by Means of Integral Transport Methods}
author = {Carlvik, I}
abstractNote = {Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.}
place = {Sweden}
year = {1967}
month = {May}
}
title = {Calculations of Neutron Flux Distributions by Means of Integral Transport Methods}
author = {Carlvik, I}
abstractNote = {Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.}
place = {Sweden}
year = {1967}
month = {May}
}