The capability of SN methodology to accurately predict the neutronics behavior of a compact, light water‐moderated reactor is examined. This includes examining the effects of cross‐section modeling and the choice of spatial and angular representation. The isothermal temperature coefficient in the range of 293 K to 355 K is analyzed, as well as the radial fission density profile across the central fuel element. Measured data from a series of critical experiments are used for these analyses.
SN transport analyses of critical reactor experiments for the SNTP program
10th SymposiumùSpace Nuclear Power and Propulsion ; 1993 ; Albuquerque, NM, USA
AIP Conference Proceedings ; 271 , 2 ; 811-815
15.01.1993
5 pages
Aufsatz (Konferenz)
Elektronische Ressource
Englisch
TEMPERATURE COEFFICIENT , POWER DENSITY , CRITICALITY EXPERIMENTS , ANGULAR DISTRIBUTION , CRITICALITY , A CODES , ENDF CROSS SECTIONS , MULTIGROUP THEORY , N CODES , WATER MODERATED REACTORS , SPACE PROPULSION REACTORS , SPACE DEPENDENCE , NEUTRON TRANSPORT THEORY , CROSS SECTIONS , SN METHODOLOGY , REACTIVITY , NEUTRON FLUX
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