• Korean Chemical Engineering Research
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• v.48 no.2
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• pp.185-192
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• 2010

The chemical-looping combustion(CLC) has advantages of no energy loss for separation of $CO_2$ without $NO_x$ formation. This CLC system consists of oxidation and reduction reactors where metal oxides particles are circulating through these two reactors. In the present study, the reaction kinetic equations of iron oxide oxygen carriers supported on bentonite have been determined by the shrinking core model. Based on the reactivity data, design values of solid circulation rate and solids inventory were determined for the rector. Two types of interconnected fluidized bed systems were designed for CLC application, one system consists of a riser and a bubbling fluidized bed, and the other one has a riser and two bubbling fluidized beds. Solid circulation rates were varied to about $30kg/m^2s$ by aeration into a loop-seal. Solid circulation rate increases with increasing aeration velocity and it increases further with an auxiliary gas flow into the loop-seal. As solid circulation rate is increased, solid hold up in the riser increases. A typical gas leakage from the riser to the fluidized bed is found to be less than 1%.

• Transactions of the Korean hydrogen and new energy society
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• v.28 no.4
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• pp.384-391
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• 2017

Effects of operating variables on solid circulation rate were measured and discussed using two-interconnected circulating fluidized bed system at ambient temperature and pressure. OCN 706-1100 particles were used as oxygen carrier. The measured solid circulation rates increased as the lower loop seal gas flow rates and the solid height in the fuel reactor increased. Suitable operating conditions to avoid choking of the air reactor were confirmed. Continuous long-term operations of steady-state solid circulation were also demonstrated at two different conditions based on the operating window.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.4
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• pp.435-442
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• 2004

This paper presents a hermetic MEMS on-chip package bonded by a closed-loop AuSn solder-line. We design three different package specimens, including a substrate heated specimen without interconnection-line (SHX), a substrate heated specimen with interconnection-line (SHI) and a locally heated specimen with interconnection-line (LHI). Pressurized helium leak test has been carried out for hermetic seal evaluation in addition to the critical pressure test for bonding strength measurement. Substrate heating method (SHX, SHI) requires the bonding time of 40min. at 400min, while local heating method (LHI) requires 4 min. at the heating power of 6.76W. In the hermetic seal test. SHX, SHI and LHI show the leak rates of 5.4$\pm$6.7${\times}$$^{-10}$ mbar-l/s, 13.5$\pm$9.8${\times}$$^{-10}$ mbar-l/s and 18.5$\pm$9.9${\times}$$^{-10}$ mbar-l/s, respectively, for an identical package chamber volume of 6.89$\pm$0.2${\times}$$^{-10}$. In the critical pressure test, no fracture is found in the bonded specimens up to the applied pressure of 1$\pm$0.1MPa, resulting in the minimum bonding strength of 3.53$\pm$0.07MPa. We find that the present on-chip packaging using a closed AuSn solder-line shows strong potential for hermetic MEMS packaging with interconnection-line due to the hermetic seal performance and the shorter bonding time for mass production.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• Proceedings of the Korean Geotechical Society Conference
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• 2009.03a
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• pp.179-187
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• 2009

Bentonite-based grouting has been popularly used to seal a borehole installed for a closed-loop vertical ground heat exchanger in a geothermal heat pump system (GHP) because of its high swelling potential and low hydraulic conductivity. The bentonite-based grout, however, has relatively lower thermal conductivity than that of ground formation. Accordingly, it is common to add some additives such as silica sand to the bentonite-based grout for enhancing thermal performance. In this study, graphite is adapted to substitute silica sand as an addictive because graphite has very high thermal conductivity. The effect of graphite on the thermal conductivity of bentonite-based grouts has been quantitatively evaluated for seven bentonite grouts from different product sources. In addition, comparisons of viscosity between applications of graphite and silica sand as additives has been carried out. In conclusion, using graphite has thermal conductivity about three times higher than that of silica sand.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.403-410
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• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.719-726
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• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.17-28
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• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.