Neutronics analysis of Dalat Research Reactor
Creators
- 1. Reactor Physics and Engineering Department, Reactor Center, Nuclear Research Inst., Dalat (Viet Nam)
Description
Many neutronics codes have been used to calculate for Dalat Research Reactor (DRR) from 1983 (the first critical of DRR in December, 1983). The purposes of all calculations are to know exactly many important parameters related to Reactor Physics and Neutron Physics in reactor core. The results from calculation play important role in core and fuel management for DRR. Especially basing on the results we can predict about fuel cycle, fuel burn up distribution and plan for using optimize remain fresh fuel assemblies of DRR. By using system neutronics code including transport codes, diffusion codes and Mote Carlo code, many characteristics of fuel assemblies and other parameters of whole core were received such as main features of VVR-M2 fuel assembly type, multiplication factor, neutron flux distribution, power distribution, burn up distribution, excess reactivity, control rods worth, neutron spectrum, temperature reactivity coefficient ect. In the paper, brief description all computer codes to being used in DRR and the calculation results from the codes above are presented. (author)
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Additional details
Publishing Information
- Imprint Title
- Proceedings of the FNCA 2003 workshop on the utilization of research reactors (Contract research)
- Imprint Pagination
- 370 p.
- Journal Page Range
- p. 141-154
- Report number
- JAEA-Conf--2006-001
Conference
- Title
- FNCA 2003 workshop on the utilization of research reactors
- Dates
- 12-16 Jan 2004
- Place
- Dalat (Viet Nam); Jakarta (Indonesia); Serpong (Indonesia)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 37103216
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BURNUP; COMPUTER CODES; CONTROL ROD WORTHS; FUEL ASSEMBLIES; FUEL MANAGEMENT; MONTE CARLO METHOD; NEUTRON FLUX; NEUTRON SPECTRA; POOL TYPE REACTORS; POWER DISTRIBUTION; REACTOR CORES; TEMPERATURE COEFFICIENT; TRIGA-2-DALAT REACTOR
- Descriptors DEC
- CALCULATION METHODS; ENRICHED URANIUM REACTORS; HOMOGENEOUS REACTORS; HYDRIDE MODERATED REACTORS; IRRADIATION REACTORS; ISOTOPE PRODUCTION REACTORS; MANAGEMENT; NUCLEAR MATERIALS MANAGEMENT; RADIATION FLUX; REACTIVITY COEFFICIENTS; REACTOR COMPONENTS; REACTORS; RESEARCH AND TEST REACTORS; SOLID HOMOGENEOUS REACTORS; SPECTRA; THERMAL REACTORS; TRIGA TYPE REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 13 refs., 11 figs., 6 tabs.