In PWR, a Departure from Nucleate Boiling (DNB) is one of criteria for the thermal-hydraulic design and safety analysis. A sub-channel analysis code calculates local coolant conditions to evaluate the PWR safety margins such as a DNB Ratio (DNBR). Mitsubishi Heavy Industries, LTD (MHI) has developed Mitsubishi Three Dimensional Drift flux Code for Analysis of Core Two-Phase Flow (MIDAC) that began the development since the 1990s which is a sub-channel analysis code for DNBR and Peak Cladding Temperature (PCT) evaluations. The code design is based on a drift flux model for the two-phase flow and a radial heat conduction model for the fuel rod temperatures. MIDAC has been verified by comparisons with exact solutions and other codes, and validated by comparisons with test data based on a Phenomena Identification and Ranking Table (PIRT) under the core thermal-hydraulic design and safety analysis conditions. As a result, MHI confirmed the applicability of MIDAC to PWR conditions in the thermal-hydraulic design and Non-LOCA.
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2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5784-7
PROCEEDINGS PAPER
Development of a Sub-Channel Analysis Code “MIDAC” for Core Thermal-Hydraulic Design and Safety Analysis in PWR Plants Available to Purchase
Tadakatsu Yodo,
Tadakatsu Yodo
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
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Naohiro Takeda,
Naohiro Takeda
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
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Naoko Iida,
Naoko Iida
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
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Motoko Kawachi
Motoko Kawachi
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
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Tadakatsu Yodo
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
Naohiro Takeda
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
Naoko Iida
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
Motoko Kawachi
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
Paper No:
ICONE25-67657, V006T08A107; 9 pages
Published Online:
October 17, 2017
Citation
Yodo, T, Takeda, N, Iida, N, & Kawachi, M. "Development of a Sub-Channel Analysis Code “MIDAC” for Core Thermal-Hydraulic Design and Safety Analysis in PWR Plants." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 6: Thermal-Hydraulics. Shanghai, China. July 2–6, 2017. V006T08A107. ASME. https://doi.org/10.1115/ICONE25-67657
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