Reactor subchannel analysis -- Electric Power Research Institute perspective
- Electric Power Research Inst., Palo Alto, CA (United States)
One of the basic objectives of subchannel flow simulation and analysis effort sponsored by the Electric Power Research Institute was the development of a computer code for subchannel analysis and its verification and validation for applications to reactor thermal margin evaluation under steady and transient conditions. A historical perspective is given of the development of specifications for a reactor core subchannel thermal-hydraulics analysis code for utility applications in the evaluation of reactor safety limits during normal operation and accident scenarios. The subchannel analysis capabilities of the VIPRE-01 code based on the homogeneous equilibrium with the algebraic slip model of two-phase flow are presented. The code, which received a safety evaluation report from the US Nuclear Regulatory Commission in 1986, is in wide use in the utility industry for fuel reload safety analysis, critical heat flux correlation development and testing, thermal margin analysis, and core thermal-hydraulic analysis. A considerable amount of work has been done during the past few years on the development of VIPRE-02, an advanced subchannel analysis code based on the two-fluid model of two-phase flow capable of simulating reactor cores, vessels, and internal structures. The functional specifications, development of VIPRE-02, and current applications for VIPRE-02, such as boiling water reactor mixed fuel core evaluation, are also discussed. Code is also used for PWR`s.
- OSTI ID:
- 170258
- Journal Information:
- Nuclear Technology, Vol. 112, Issue 3; Other Information: PBD: Dec 1995
- Country of Publication:
- United States
- Language:
- English
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