Published February 1994
| Version v1
Journal article
Two-group macroscopic cross-section data base for the prototype Miniature Neutron Source Reactor
Description
A two-group macroscopic cross-section data base has been generated using a simplified model for the Miniature Neutron Source Reactor (MNSR). The correction term to the basic macroscopic cross-sections due to varying effects of temperature along the axis of the fuel and burnup were expressed in the form of poly-nomials. The error analysis that the database is accurate enough to be used for neutronic calculations for the reactor. (author). 9 refs.; 2 tabs.; 19 figs
Additional details
Publishing Information
- Journal Title
- Journal of the Kumasi University of Science and Technology
- Journal Volume
- 14
- Journal Issue
- 1
- Journal Page Range
- p. 40-47.
- ISSN
- 0855-0395
- CODEN
- JKUTDP
INIS
- Country of Publication
- Ghana
- Country of Input or Organization
- Ghana
- INIS RN
- 25073836
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- BURNUP; COMPUTERIZED SIMULATION; GROUP CONSTANTS; HOMOGENIZATION METHODS; MULTIGROUP THEORY; NEUTRON DIFFUSION EQUATION; REACTOR CORES
- Descriptors DEC
- CALCULATION METHODS; CROSS SECTIONS; REACTOR COMPONENTS; SIMULATION; TRANSPORT THEORY