Published 1997
| Version v1
Journal article
An experience with the use of MCNP code and ENDF/B data libraries to criticality calculations in the case of WWER fuel assemblies
Description
The criticality problem of the neutron transport equation was solved for various configurations of the fuel rod lattices by using the MCNP code. A continuous representation of the cross-sections was employed and all computations were performed multiply using data based on the ENDF/B-4, ENDF/B-5, and ENDF/B-6 libraries, respectively. The data library used was found to affect appreciably the resulting multiplication factor value. 1 tab., 5 figs., 8 refs
Additional details
Publishing Information
- Journal Title
- Nucleon
- Journal Issue
- no.1
- Journal Page Range
- p. 7-10.
- ISSN
- 1210-2660
- CODEN
- NLEQEM
INIS
- Country of Publication
- Czech Republic
- Country of Input or Organization
- Czech Republic
- INIS RN
- 28053741
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S99: GENERAL AND MISCELLANEOUS;
- Descriptors DEI
- CONFIGURATION; CRITICALITY; CROSS SECTIONS; FUEL RODS; M CODES; MONTE CARLO METHOD; MULTIPLICATION FACTORS; NUCLEAR DATA COLLECTIONS; REACTOR LATTICES; WWER TYPE REACTORS
- Descriptors DEC
- CALCULATION METHODS; COMPUTER CODES; ENRICHED URANIUM REACTORS; FUEL ELEMENTS; POWER REACTORS; PWR TYPE REACTORS; REACTOR COMPONENTS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS