Korea Institute of Nuclear Safety
Kukhee Lim, Yong Jin Cho, and Jung Jae Lee
Technical Meeting on Phenomenology and Technologies Relevant to In-Vessel Melt Retention and Ex-Vessel Corium Cooling 17–21 October 2016, Shanghai, China
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• 25 Operable Power Reactors
• 3 PWRs under Construction
• 8 Additional APR1400 or APR+ Units by 2029 (planned)
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• IVR-ERVC is selected as an Severe Accident Management Strategy for APR1400.
• Design Requirements
• Implementation
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• ERVC and CFS(Cavity Flooding System)
• International cooperation at early 2000s (In-vessel Retention Strategy for High Power Reactors*)
• ERVC experiment
• Core degradation analysis
• Lower head penetration experiment
• Code development
8 *INEEL/EXT-04-02561, 2005
• Project : In-vessel Retention Strategy for High Power Reactors*
• Participants : INEEL, SNU, PSU, KAERI (from 2002 to 2004)
9 *INEEL/EXT-04-02561, 2005
Initial code analyses
ERVC experiments Code analyses using modified design
• 3-D hemisphere (D = 0.11m), vessel wall heating, natural circulation
• Nucleate boiling and CHF experiment with enhanced insulation structure and coated vessel surface (considering ICI guide tubes)
• Max. CHF
• Limitations
10 * J. Yang, Ph. D Dissertation, PSU, 2005
• Heat transfer correlations (implemented in MAAP5)
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Nucleate boiling correlation
ℎ = 𝑓(∆𝑇, 𝜃)
* J. Yang, Ph. D Dissertation, PSU, 2005
Critical heat flux correlation
ℎ = 𝑓(𝜃)
• Air-water two-phase (non-
heating) ex-vessel cooling
channel experiment
• ½ scale geometry (D = 2.610 m)
• To observe and evaluate the two-
phase natural circulation phenomena
through the gap such as
• Limitations
12 *INEEL/EXT-04-02561, 2005
• 1/10 scale 3-D hemisphere, Cu ex-vessel surface,
vessel wall heating, forced circulation
• Different heat flux input at oxidic and metallic
molten pool region
• CHF : 1.31-1.56 MW/m2
(depending on mass flow)
• Limitations
13 * S. W. Noh, et. al., NED, Vol. 258, p. 116-129, 2013
• Independent calculations of INEEL and KAERI using
SCDAP/RELAP5-3Dⓒ and SCDAP/RELAP5/Mod3.3
14 *INEEL/EXT-04-02561, 2005
• INEEL VESTA
• KAERI LILAC-LP
15 *INEEL/EXT-04-02561, 2005
• Purpose
• 1st campaign
• 2nd campaign
• Limitations
16 * Y. H. Jeong et. al., Nuclear Technology, Vol. 152, p. 162-169, 2005
• Development of CHF model for a downward-facing curved surface
17 * H. M. Park, Ph. D dissertation, KAIST, 2014
• MAAP5 analysis of IVR strategy for Shinkori 3&4 NPPs
(for TLOFW, STGR, SBLOCA, and LBLOCA scenarios)
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18 * KEPCO E&C, Shinkori 3&4 IVR-ERVC analysis report(Rev. 1), 2014
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19 * KEPCO E&C, Shinkori 3&4 IVR-ERVC analysis report(Rev. 1), 2014
• KAERI VESTA-S
20 * J. H. Song, Presentation at PLINUS-2 International Seminar Marseille, France May 16th, 2014
• Development of CSPACE (KAERI, since 2014)
21 * D. G. Son, et. al., IVR workshop, Aix-en-Provence, France, 6-7th June 2016
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22 * K. H. Lim. et. al., CSARP & MCAP meeting, Bethesda, MD, Sep. 13-16, 2016
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• Uncertainty analysis using lumped parameter method (with POSTECH)
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0.0 0.5 1.0 1.5 2.0 2.50.0
0.2
0.4
0.6
0.8
1.0
SBLB
ULPU
Thermal failure
criterion
Top of
the metal layer
Top of
the oxide layer
Cum
ula
tive P
robabili
ty
Heat flux ratio (q"/q"CHF
)
Bottom of
the oxide layer
0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.50.0
0.2
0.4
0.6
0.8
1.0
K-E & M
mini-ACOPO
ACOPO
BALI
Top of the
metal layer
Top of the
oxide layer
Cu
mu
lative
Pro
ba
bili
ty
Heat flux (MW/m2)
Bottom of
the oxide layer
• Lower head integrity
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• Experimental approach
• Analytical approach
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