Korea Institute of Nuclear Safety - International Atomic ... · PDF fileKorea Institute of...

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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