A cross-platform multi-physics reactor transient simulation framework through a SALOME-Backbone joint platform
Creators
- 1. Canadian Nuclear Labs., Chalk River, ON (Canada)
- 2. Univ. of Toronto, Toronto, ON (Canada)
Description
Predicting nuclear reactor behaviour during transient conditions can be enhanced through numerical multi-physics simulation framework involving neutron transport, thermal hydraulics, and fuel behaviour codes to improve nuclear reactor safety analysis. Two tools are used in this work to accomplish the coupling of these codes. One is SALOME, an open source software that provides a generic platform for Pre- and Post-Processing of numerical simulations. Another one is Backbone, developed at the Canadian Nuclear Laboratories, which has the capability of coupling multi-physics codes under different operating systems. Also, Backbone is capable of synchronizing various scientific codes through a Time Step Controller (TSC), which permits appropriate and efficient data exchange at the time-step level. This paper studies a SALOME-Backbone joint platform for code coupling with comparisons made to a Large Break Loss of Coolant Accident (LB-LOCA) experiment performed in a five channel per pass configuration at the CNL Whiteshell RD-14M facility. (author)
Additional details
Publishing Information
- Publisher
- Canadian Nuclear Society
- Imprint Place
- Toronto, Ontario (Canada)
- ISBN
- 978-1-926773-24-7
- Imprint Title
- CANDU fuel : evolution towards optimal performance. 13th international conference on CANDU fuel
- Imprint Pagination
- 66.6 Megabytes
- Journal Page Range
- [14 p.]
Conference
- Title
- 13. International conference on CANDU fuel
- Dates
- 15-18 Aug 2016
- Place
- Kingston, Ontario (Canada)
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 50005396
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- COMPUTER CODES; LBLOCA; NEUTRON TRANSPORT; REACTOR ACCIDENT SIMULATION; SAFETY ANALYSIS; THERMAL HYDRAULICS
- Descriptors DEC
- ACCIDENTS; FLUID MECHANICS; HYDRAULICS; LOSS OF COOLANT; MECHANICS; NEUTRAL-PARTICLE TRANSPORT; RADIATION TRANSPORT; REACTOR ACCIDENTS; SIMULATION
Optional Information
- Notes
- 16 refs., 1 tab., 6 figs.