Thermal-hydraulic interfacing code modules for CANDU reactors
Conference
·
OSTI ID:544402
- Ontario Hydro Nuclear, Toronto (Canada); and others
The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.
- Research Organization:
- Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; Nuclear Energy Agency, 75 - Paris (France); SCIENTECH, Inc., Boise, ID (United States)
- OSTI ID:
- 544402
- Report Number(s):
- NUREG/CP--0159; NEA/CSNI/R--(97)4; CONF-961192--; ON: TI97008508
- Country of Publication:
- United States
- Language:
- English
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