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Development of a Sub-Channel Analysis Code “MIDAC” for Core Thermal-Hydraulic Design and Safety Analysis in PWR Plants

[+] Author Affiliations
Tadakatsu Yodo, Naohiro Takeda, Naoko Iida, Motoko Kawachi

Mitsubishi Heavy Industries, Ltd., Kobe, Japan

Paper No. ICONE25-67657, pp. V006T08A107; 9 pages
doi:10.1115/ICONE25-67657
From:
  • 2017 25th International Conference on Nuclear Engineering
  • Volume 6: Thermal-Hydraulics
  • Shanghai, China, July 2–6, 2017
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5784-7
  • Copyright © 2017 by ASME

abstract

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.

Copyright © 2017 by ASME

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