Published 1983
| Version v1
Book
An alternate approach to neutronics analysis of CANDU reactors
Description
An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Uptil now CANDU reactors have been analysed by the methods developed at AECL and CGE using semi-empirical techniques. Relying on multi-group transport codes GAM and GATHER in combination with diffusion code CITATION a package of codes is established with the object of using it for survey as well as production purposes. A typical 600 MW(e) CANDU reactor has been analysed and the results so obtained are quite reasonable
Additional details
Publishing Information
- Publisher
- American Nuclear Society.
- Imprint Place
- LaGrange Park, IL (USA)
- ISBN
- 0-919784-02-X
- Imprint Title
- Proceedings of the international conference on numerical methods in nuclear engineering. Volume 1
- Journal Page Range
- p. 690-697.
Conference
- Title
- International conference on numerical methods in nuclear engineering.
- Dates
- 6-9 Sep 1983.
- Place
- Montreal (Canada).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 17051822
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- C CODES; CANDU TYPE REACTORS; COMPUTER CALCULATIONS; DIFFUSION; G CODES; NEUTRON TRANSPORT; NEUTRON TRANSPORT THEORY
- Descriptors DEC
- COMPUTER CODES; HEAVY WATER MODERATED REACTORS; NEUTRAL-PARTICLE TRANSPORT; POWER REACTORS; PRESSURE TUBE REACTORS; RADIATION TRANSPORT; REACTORS; THERMAL REACTORS; TRANSPORT THEORY