Experimental Study on Critical Power in a Hemispherical Narrow Gap

Park, Rae-Joon; Ha, Kwang Soon; Kim, Sang-Baik; Kim, Hee‐Dong; Jeong, Jin Hee

doi:10.13182/nt02-a3338

Cited by 11 publications

(11 citation statements)

References 10 publications

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“…Figures 2 and 3 indicate that quenching tests give almost the same heat transfer rate and CCFL constant as the quasi-steady state CHF tests when the flow patterns are the same in both tests. (7) and (10) by Park et al 8) and Eqs. In the test by Chun et al, a heater rod with the outer diameter of 9.54 mm and the heated length of 1,842 mm is inserted in a tube with the inner diameter of 19.4 mm, the top end of the annulus is open and its bottom end is closed.…”

Section: Chf and Ccfl In Narrow Gapmentioning

confidence: 90%

“…Okano et al 12) obtained the empirical constant C K from the measured q CHF;A of existing experiments [2][3][4][5][6][7][8][9][10] and Eq. (13) is used in place of the hydraulic diameter D h in the Wallis correlation.…”

Section: Chf and Ccfl In Narrow Gapmentioning

confidence: 99%

“…Many experiments on gap cooling have been performed [2][3][4][5][6][7][8][9] and heat transfer models have been proposed as gap cooling critical heat flux (CHF). However, these experiments were performed under the quasi-steady state and low superheat conditions.…”

Section: Introductionmentioning

confidence: 99%

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Modeling and Validation of In-Vessel Debris Cooling during LWR Severe Accident

Okano¹,

Nagae²,

Murase³

2005

Journal of Nuclear Science and Technology

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The authors proposed the heat transfer models to evaluate in-vessel debris cooling during a severe accident in a light water reactor (LWR), where the heat flux on the vessel inner surface was restricted by countercurrent flow limitation (CCFL) at the top end of the narrow gap between the overheated core debris and the reactor pressure vessel (RPV) or the local boiling heat flux, and derived correlations for CCFL and boiling heat fluxes using the existing data. In this paper, we improved the correlation of nucleate boiling heat fluxes using quenching data at the pressure of 0.1 MPa and the ALPHA test data at 1.3 MPa. Using the correlations, we calculated transient vessel temperatures in the LAVA experiment at 1.7 MPa. During the heating process of the vessel, the calculated average temperature was determined by CCFL and agreed well with the average of the measured temperatures. During its cooling process, the calculated local cooling rate of the vessel without the CCFL correlation was greatly affected by the correlation of nucleate boiling heat fluxes and the calculated results using the improved correlation agreed well with the measured values. As a result, applicability of the correlations for CCFL and boiling heat fluxes to in-vessel cooling was validated.

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Section: Chf and Ccfl In Narrow Gapmentioning

confidence: 90%

Section: Chf and Ccfl In Narrow Gapmentioning

confidence: 99%

See 1 more Smart Citation

Modeling and Validation of In-Vessel Debris Cooling during LWR Severe Accident

Okano¹,

Nagae²,

Murase³

2005

Journal of Nuclear Science and Technology

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“…The CCFL correlation below was formulated from the experiments by Park et al for a vertical annulus with an outer diameter of 0.5 m, 22) which were large scale experiments in terms of a peripheral length of narrow gap,…”

Section: Critical Heat Flux In Narrow Gapmentioning

confidence: 99%

“…The experiments had geometries of vertical annuli (Koizumi et al), 23) vertical rectangular plates (CTF experiments) 24) and hemispherical gap (CHF experiments). 22) These experiments covered an ambient pressure from 0.1 to 8.0 MPa, a gap width from 0.5 to 5.0 mm and Bond number (Bo) from 0.16 to 98.0 as shown in Table 1.…”

Section: Critical Heat Flux In Narrow Gapmentioning

confidence: 99%

Modeling for Evaluation of Debris Coolability in Lower Plenum of Reactor Pressure Vessel

Maruyama

Moriyama

Nakamura

et al. 2003

Journal of Nuclear Science and Technology

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Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code.

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Experimental investigations of the quenching phenomena for hemispherical downward facing convex surfaces with narrow gaps

Park

Kim

et al. 2007

International Communications in Heat and Mass Transfer

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Experimental Study on Critical Power in a Hemispherical Narrow Gap

Cited by 11 publications

References 10 publications

Modeling and Validation of In-Vessel Debris Cooling during LWR Severe Accident

Modeling and Validation of In-Vessel Debris Cooling during LWR Severe Accident

Modeling for Evaluation of Debris Coolability in Lower Plenum of Reactor Pressure Vessel

Experimental investigations of the quenching phenomena for hemispherical downward facing convex surfaces with narrow gaps

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