VIPRE-A; Reactor core thermal-hydraulics analysis code for utility applications

Journal Article · Sun Nov 01 04:00:00 EST 1992 · Nuclear Technology; (United States)

OSTI ID:7016120

Srikantiah, G S ^[1]

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1. Electric Power Research Inst., Palo Alto, CA (United States)

In this paper the development, validation, and applications of the VIPRE code are presented. 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 is traced. The capabilities of the VIPRE-01 code based on a homogeneous equilibrium model of two-phase flow with algebraic slip are presented along with a discussion of the extensive verification and validation of the code. Utility applications of the code, which received a safety evaluation report from the U.S. Nuclear Regulatory Commission in 1986, in the areas of fuel reload safety analysis, critical heat flux correlation development and testing, thermal margin analysis, and core thermal-hydraulic analysis are presented. The functional specifications and the development of VIPRE-02, an advanced version of the code based on a two-fluid model of two-phase flow that is capable of simulating the reactor core, vessel, and internal structure, are also described. A discussion of the developing applications for VIPRE-02, such as boiling water reactor instability analysis and pressurized water reactor steamline break analysis, is given with some initial results.

• OSTI ID: 7016120
• Journal Information: Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 100; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

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