• Korean Chemical Engineering Research
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• 제48권2호
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• pp.185-192
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• 2010

매체 순환식 연소는 연소 공정 자체에서 질소 산화물 생성이나 부가적인 에너지 소비 없이 이산화탄소 분리가 이루어지는 신공정이다. 이 공정은 금속 산화물 입자가 두 개의 반응기를 순환하며 산화와 환원을 거치는 과정으로 구성되어 있다. 이 연구에서는 bentonite에 담지된 산화철 산소 공여 입자의 반응 속도 식을 shrinking core 모델을 통하여 수립하였다. 반응성 결과를 바탕으로 반응기 설계 기준인 고체 순환량과 입자 충전량을 도출하였다. 매체 순환식 연소 공정의 적용을 위하여 두 가지 형태의 연결된 유동층 즉, 상승관과 기포 유동층이 각각 한 개씩인 형태, 상승관 한 개와 기포 유동층이 두 개로 구성된 형태로 시스템을 설계하였다. 고체 순환량은 loop-seal을 통하여 $30kg/m^2s$ 정도까지 변화시켰다. 고체 순환량은 loop-seal의 기체 주입량이 증가할수록 증가하였으며 보조 기체를 주입하면 그 양이 더 증대되었다. 고체 순환량이 증가함에 따라 상승관 내부의 고체량은 증가하였다. 상승관으로부터 다른 반응기로의 기체 누출량은 1% 미만의 수준이었다.

• 한국수소및신에너지학회논문집
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• 제28권4호
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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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• 제44권7호
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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.

• 대한기계학회논문집A
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• 제28권4호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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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.

• 한국지반공학회:학술대회논문집
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• 한국지반공학회 2009년도 춘계 학술발표회
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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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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• 제44권7호
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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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• 제41권5호
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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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• 제31권1호
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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.