Thermal-hydraulic interfacing code modules for CANDU reactors

Conference · Tue Jul 01 00:00:00 EDT 1997

OSTI ID:544402

Liu, W S; Gold, M; Sills, H ^[1]

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1. 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: US Nuclear Regulatory Commission (NRC), 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; TRN: 98:000240
• Resource Relation: Conference: Organization for Economic Co-Operation and Development (OECD)/Committee on the Safety of Nuclear Installations (CSNI) workshop on transient thermal-hydraulic codes requirements, Annapolis, MD (United States), 5-8 Nov 1996; Other Information: PBD: Jul 1997; Related Information: Is Part Of Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements; Ebert, D.; PB: 824 p.
• Country of Publication: United States
• Language: English

Related Subjects

21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
22 NUCLEAR REACTOR TECHNOLOGY
99 MATHEMATICS
COMPUTERS
INFORMATION SCIENCE
MANAGEMENT
LAW
MISCELLANEOUS
CANDU TYPE REACTORS
HYDRAULICS
SAFETY ANALYSIS
COMPUTER CODES
MODIFICATIONS
COMPUTERIZED SIMULATION