• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Nuclear industry
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• v.5 no.9 s.31
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• pp.2-5
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• 1985

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.227.2-227
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.141-141
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• 1998

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.665-674
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• 2006

Three-dimensional neutron flux and fluence of Kori Unit 3 were evaluated using the synthesis technique described in Regulatory Guide 1.190 for all reactor geometry. For this purpose DORT neutron transport calculations from Cycle 1 to Cycle 15 were performed using BUGLE-96 cross-section library. The calculated flux and fluence were validated by comparing the calculated reaction rates to the measurement data from the dosimetry sensor set of the $5^{th}$ surveillance capsule withdrawn at the end of cycle 15 of Kori Unit 3. And then the best estimation of the neutron exposures for the reactor vessel beltline region was performed using the least square evaluation. These results can be used in the assessment of the state of embrittlement of Kori Unit 3 pressure vessel.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.247-247
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.309-314
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• 1996

Feasibility study has been accomplished to evaluate the effectiveness of the in-vessel core debris cooling through lower cavity flooding using two dimensional finite difference scheme. The volume of cerium pool and decay power rate generated in corium pool were evaluated as important parameters to the temperature distribution on the reactor vessel lower head through previous works. In this study, the corium volume based on the System 80＋ core structure and time dependent decay power rate are considered for feasibility evaluation. In addition, preliminary plans for the in-vessel core debris cooling through lower cavity flooding as severe accident management strategy, i.e. flooding timing, method and capacity, are suggested based on the result of the numerical study, international tendency related to in-vessel core debris cooling through lower cavity flooding.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.190-190
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• 1999

CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.535-542
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• 2012

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.290.2-290
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• 2002