Scalable three-dimensional thermal-hydraulic best-estimate code BAGIRA

Conference · Sun Jul 01 04:00:00 EDT 2012

OSTI ID:22107822

Vasenin, V. A.; Krivchikov, M. A. ^[1]; Kroshilin, V. E.; Kroshilin, A. E.; Roganov, V. A. ^[2]

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1. Moscow State Univ., Moscow (Russian Federation)
2. All-Russian Scientific Research Inst. of Nuclear Power Plants (VNIIAES), Ferganskaya 25, Moscow (Russian Federation)

The three-dimensional thermal-hydraulic best-estimate code BAGIRA for modeling of multi-phase flows was developed without any artificial physical assumptions or simplifications. The mathematical model is based on numerical approximations of exact three-dimensional equations, including effective multi-dimensional models for turbulent heat and mass transfer. With use of BAGIRA All-Russian Scientific Research Inst. of Nuclear Power Plants (VNIIAES) has developed a full-scope and analytical simulators using BAGIRA for a number of power plants with VVER-1000 and RBMK type design, which are being used in Kalinin, Kursk, Smolensk, Chernobyl, and Bilibino NPPs. The comparison of calculated and experimental results shows that BAGIRA can successfully reproduce the most important processes observed in experiments. BAGIRA is implemented in FORTRAN. It is a relatively complicated code that tends to decompose task by aspects. Such a style is welcoming for extensions, which can be added without code redesign. We would like to present an aspect-oriented mix-in approach for BAGIRA code extension. It allows to make it scalable in number of directions leaving original code base untouched. It is possible to add new effects/units, and even to produce a supercomputer version of the code. The last is a key point today due to availability of low-cost compact supercomputers, which makes building compact NPP simulators possible. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22107822
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICS AND COMPUTING
APPROXIMATIONS
COMPARATIVE EVALUATIONS
DESIGN
HEAT TRANSFER
MASS TRANSFER
MATHEMATICAL MODELS
MULTIPHASE FLOW
NUCLEAR POWER PLANTS
REACTOR SIMULATORS
SIMULATION
SUPERCOMPUTERS
THERMAL HYDRAULICS
THREE-DIMENSIONAL CALCULATIONS
WWER TYPE REACTORS