• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.235.2-235
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• 2001

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• Journal of the Korean Society of Fisheries and Ocean Technology
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• v.34 no.2
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• pp.200-211
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• 1998

Heat transfer performance improvement by fin and groovs is studied for condensation of R-11 on integral-fin tubes. Eight tubes with trapczodially shaped integral-fins having fin density from 748 to 1654fpm(fin per meter) and 10, 30 grooves are tested. A plain tube having the same diameter as the finned tubes is also used for comparison. R-11 condensates at saturation state of 32 $^{\circ}C$ on the outside tube surface coded by inside water flow. All of test data are taken at steady state. The heat transfer loop is used for testing singe long tubes and cooling is pumped from a storage tank through filters and folwmeters to the horizontal test section where it is heated by steam condensing on the outside of the tubes. The pressure drop across the test section is measured by menas pressure gauge and manometer. The results obtained in this study is as follows : 1. Based on inside diameter and nominal inside area, overall heat transfer coefficients of finned tube are enhanced up to 1.6 ~ 3.7 times that of a plain tube at a constant Reynolds number. 2. Friction factors are up to 1.6 ~ 2.1 times those of plain tubes. 3. The constant pumping power ratio for the low integral-fin tubes increase directly with the effective area to the nominal area ratio, and with the effective area diameter ratio. 4. A tube having a fin density of 1299fpm and 30 grooves has the best heat transfer performance.

• Proceedings of the IEEK Conference
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• 2009.05a
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• pp.326-328
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• 2009

we an autotuning algorithm for PI controller with unknown plant. The proposed algorithm uses a saturation function and time delay element as a test signal. Since the integral element of PI controller reduces a phase margin and amplitude margin in the closed loop system, the closed loop system could be resulted in unstable with PI controller, To avoid unstable in the closed loop system with PI controller, the proposed algorithm identifies one point information in the 3rd quadrant of Nyquist plot with a time delay element. The proposed method improves an accuracy of one point identified information with one saturation function.

• Journal of the Korean Society of Fisheries and Ocean Technology
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• v.34 no.2
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• pp.212-222
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• 1998

Heat transfer performance improvement by fin and groovs is studied for condensation of R-11 on integral-fin tubes. Eight tubes with trapczodially shaped integral-fins having fin density from 748 to 1654fpm(fin per meter) and 10, 30 grooves are tested. A plain tube having the same diameter as the finned tubes is also used for comparison. R-11 condensates at saturation state of 32 $^{\circ}C$ on the outside tube surface coded by inside water flow. All of test data are taken at steady state. The heat transfer loop is used for testing singe long tubes and cooling is pumped from a storage tank through filters and folwmeters to the horizontal test section where it is heated by steam condensing on the outside of the tubes. The pressure drop across the test section is measured by menas pressure gauge and manometer. The results obtained in this study is as follows : 1. Based on inside diameter and nominal inside area, overall heat transfer coefficients of finned tube are enhanced up to 1.6 ~ 3.7 times that of a plain tube at a constant Reynolds number. 2. Friction factors are up to 1.6 ~ 2.1 times those of plain tubes. 3. The constant pumping power ratio for the low integral-fin tubes increase directly with the effective area to the nominal area ratio, and with the effective area diameter ratio. 4. A tube having a fin density of 1299fpm and 30 grooves has the best heat transfer performance.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.25-44
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• 2011

KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basis accidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed and successfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from the ATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. For the first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 was selected by considering its technical importance and by incorporating comments from participants. Twelve domestic organizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. This ATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participants prior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was also used by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison results between the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations for code users as well as for developers are suggested.

• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.

• International Journal of Control, Automation, and Systems
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• v.6 no.4
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• pp.515-525
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• 2008

In this paper, the problem of delay-dependent stabilization for singular systems with multiple internal and external incommensurate constant point delays is investigated. The condition when a singular system subject to point delays is regular independent of time delays is given and it can be easily test with numerical or algebraic methods. Based on Lyapunov-Krasovskii functional approach and the descriptor integral-inequality lemma, a sufficient condition for delay-dependent stability is obtained. The main idea is to design multiple memoryless state feedback control laws such that the resulting closed-loop system is regular independent of time delays, impulse free, and asymptotically stable via solving a strict linear matrix inequality (LMI) problem. An explicit expression for the desired memoryless state feedback control laws is also given. Finally, a numerical example illustrates the effectiveness and the availability for the proposed method.