Published 2002
| Version v1
Computer medium
A Monte Carlo estimate of the diffusion coefficient for a CANDU cell
Creators
- 1. Georgia Inst. of Technology, Nuclear and Radiological Engineering and Health Physics Program, George W. Woodruff School of Mechanical Engineering, Atlanta, Georgia (Canada)
- 2. Studsvik Scandpower AS, Kjeller (Norway)
Description
In this paper, the radial diffusion coefficient for a CANDU cell is calculated by using the Monte Carlo code MCNP. The calculation uses the conventional definition of the diffusion coefficient which is based on the transport cross section. The latter is estimated from an MCNP simulation of the CANDU cell with a reflective boundary condition. These MCNP-based two-group diffusion coefficients are compared to the corresponding result obtained with the transport and depletion code HELIOS. The calculations are performed for two burnup states: 0 and 4 GWd/TU. (author)
Additional details
Publishing Information
- Publisher
- Canadian Nuclear Society
- Imprint Place
- Toronto, Ontario (Canada)
- Imprint Title
- Nuclear power reactors : step into the future. 22nd Canadian Nuclear Society nuclear simulation symposium
- Imprint Pagination
- 37.9 Megabytes
- Journal Page Range
- [5 p.]
Conference
- Title
- 22. Canadian Nuclear Society nuclear simulation symposium
- Dates
- 3-5 Nov 2002
- Place
- Ottawa, Ontario (Canada)
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 37004621
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CANDU TYPE REACTORS; COMPUTER CALCULATIONS; DIFFUSION; MONTE CARLO METHOD; TRANSPORT
- Descriptors DEC
- CALCULATION METHODS; HEAVY WATER MODERATED REACTORS; POWER REACTORS; PRESSURE TUBE REACTORS; REACTORS; THERMAL REACTORS
Optional Information
- Notes
- 11 refs., 1 tab., 1 fig. Imprint:Available on Compact Disc from the Canadian Nuclear Society, Toronto, Ontario (Canada).