• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.826-833
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• 2022

A two-phase natural circulation test using SMART integral test loop (SMART-ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler-condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART-ITL was compared with those in pressurized water reactor simulators. In the SMART-ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.249.1-249
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• 2001

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.509-514
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• 2005

Main coolant pump (MCP) is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel rods and steam generators in an integral type reactor. The reactor is designed to operate under condition of 310 oC and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition in order to verify its performance and safety. In present work, a test loop to simulate real operating situation of the reactor has been designed and constructed to test MCP. And then, as a part of qualification test, canned motor functional test and pump hydraulic performance test have been carried out upon a prototype MCP. Canned motor efficiency and pump hydraulic characteristics including homologous curves and NPSH curves were obtained from the qualification test.

• Journal of Power Electronics
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• v.19 no.5
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• pp.1193-1202
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• 2019

A SMPMSM drive system is a typical nonlinear system with time-varying parameters and unmodeled dynamics. The speed outer loop and current inner loop control structures are coupled and coexist with various disturbances, which makes the speed control of SMPMSM drive systems challenging. First, an ultra-local model of a PMSM driving system is established online based on the algebraic estimation method of model-free control. Second, based on the backstepping control framework, model-free adaptive integral backstepping (MF-AIB) control is proposed. This scheme is applied to the permanent magnet synchronous motor (PMSM) drive system of an electric vehicle for the first time. The validity of the proposed control scheme is verified by system simulations and experimental results obtained from a SMPMSM drive system bench test.

• Journal of Fisheries and Marine Sciences Education
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• v.6 no.1
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• pp.58-76
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• 1994

Heat transfer and pressure drop measurements are made on low integral-fin tubes in turbulent water flow condition. The integral-fin tubes investigated in this paper are nominally 19mm in diameter. Eight tubes have been used with trapezoidally shaped integral-fins having fin density from 748 to 1654 fpm and 10, 30 grooves. Plain tube having same diameter as finned tube is also tested for comparison. Experiments are carried out using R-11 as working fluid. The refrigerant condensates at a saturation state of $30^{\circ}C$ on the outside tube surface cooled by coolant. The amount of noncondensable gases present in the test loop is reduced to a negligible value by repeated purging. For a given heat input to the boiler and given cooling water flow rate, all test data are taken on steady state. The heat transfer loop is used for testing single long tubes and cooling water is pumped from a storage tank through filters and flowmeters to the horizontal test section where it is heated by steam condensing on the outside of the tube. The pressure drop across the test section is measured by means of pressure gauge and manometer. Each tube tested is cleaned with sodium dichromate pickling solution and well rinsed with water prior to installation in the test section. The results obtained in this study is as follows : 1. Based on inside diameter and nominal inside area, heat transfer of finned tube is enhanced up to 4 times as that of a plain tube at constant Reynolds number and up to 2 times at constant pumping power. 2. Friction factors are up to 1.6~2.1 times those of plain tube. 3. At a given Reynolds number, Nusselt number decrease with increasing pitch to diameter. 4. The constant pumping power ratio for low integral-fin tubes increase directly with the effective area to the nominal area ratio, and with the effective area diameter ratio.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• International Journal of Aeronautical and Space Sciences
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• v.11 no.4
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• pp.303-312
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• 2010

This paper introduces the upper stage attitude control system of KSLV-I, which is the first space launch vehicle in Korea. The KSLV-I upper stage attitude control system consists of two electro-hydraulic actuators and a reaction control system using cold nitrogen gas. A proportional, derivative, and integral controller is designed for the electro-hydraulic thrust vectoring system, and Schmidt trigger ON/OFF controllers are designed for the reaction control system. Each attitude controller is designed to have enough stability margins. The stability and performance of KSLV-I upper stage attitude control system is verified via hardware in the loop tests. Hardware in the loop tests are accomplished for perturbed flight conditions as well as nominal flight condition. The test results show that the attitude control loop of KSLV-I upper stage is very stable and the attitude controllers perform well for all flight conditions. Attitude controllers designed in this paper have been successfully applied to the first flight of KSLV-I on August 25, 2009. The flight test results show that all attitude controllers of the KSLV-I upper stage performed well and satisfied the accuracy specifications even during abnormal flight conditions.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.353-363
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• 2023

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.