Published January 2003
| Version v1
Journal article
Monte Carlo transport and burnup calculation
Creators
Description
A 3-D multigroup P3 neutron transport Monte Carlo code MCMG-BURN is developed for coupling neutron transport with burnup. MCMG-BURN code is based on Monte Carlo code MCNP with the continuous energy cross-section and the reactor lattice code WIMS. It uses the up-front multigroup macroscopic cross-section library based on burnup for Monte Carlo calculations. The almost consistent results with the experiments have been achieved
Additional details
Identifiers
- PII
- S0306454902000440;
Publishing Information
- Journal Title
- Annals of Nuclear Energy (Oxford)
- Journal Volume
- 30
- Journal Issue
- 1
- Journal Page Range
- p. 127-132
- ISSN
- 0306-4549
- CODEN
- ANENDJ
INIS
- Country of Publication
- United Kingdom
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 33048668
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Descriptors DEI
- BURNUP; CROSS SECTIONS; M CODES; MONTE CARLO METHOD; MULTIGROUP THEORY; NEUTRON TRANSPORT THEORY; REACTOR LATTICE PARAMETERS; THREE-DIMENSIONAL CALCULATIONS; W CODES
- Descriptors DEC
- CALCULATION METHODS; COMPUTER CODES; NEUTRON TRANSPORT THEORY; TRANSPORT THEORY
Optional Information
- Copyright
- Copyright (c) 2002 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.