• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.1-6
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• 2011

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established a small-scale gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype made of Hastelloy-X to be tested in the small-scale gas loop. Results from the elastic structural analysis on the PHE prototype were reported in the previous article. In order to investigate the macroscopic structural characteristics and behavior of the PHE prototype under the test condition of the small-scale gas loop far more in detail, elastic-plastic high-temperature structural-analysis of the PHE prototype was carried out in this study.

• Korean Chemical Engineering Research
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• 제55권3호
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• pp.279-286
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• 2017

This review paper evaluates different kinds of synthesis methods for calcite precipitated calcium carbonates by using different materials. The various processing routes of calcite with different compositions are reported and the possible optimum conditions required to synthesize a desired particle sizes of calcite are predicted. This paper mainly focuses on that the calcite morphology and size of the particles by carbonation process using loop reactors. In this regard, we have investigated various parameters such as $CO_2$ flow rate, Ca $(OH)_2$ concentration, temperature, pH effect, reaction time and loop reactor mechanism with orifice diameter. The research results illustrate the formation of well-defined and pure calcite crystals with controlled crystal growth and particle size, without additives or organic solvents. The crystal growth and particle size can be controlled, and smaller sizes are obtained by decreasing the Ca $(OH)_2$ concentration and increasing the $CO_2$ flow rate at lower temperatures with suitable pH. The crystal structure of obtained calcite was characterized by using X-ray diffraction method and the morphology by scanning electron microscope (SEM). The result of x-ray diffraction recognized that the calcite phase of calcium carbonate was the dominating crystalline structure.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.327-336
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• 2007

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• Nuclear Engineering and Technology
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• 제49권5호
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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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• 제54권3호
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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.

• 한국분무공학회지
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• 제8권1호
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• pp.23-28
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• 2003

When a cold HPSI (High Pressure Safety Injection) fluid associated with a design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena may arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated collant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• 비파괴검사학회지
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• 제23권5호
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• pp.520-525
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• 2003

Fast Breeder Test Reactor (FBTR) is a 40MW (thermal) / 13.2MW (electrical), Plutonium - Uranium mixed carbide fuelled, sodium cooled, loop type nuclear reactor operating at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. Its main aim is to generate experience in operation of fast reactors and sodium systems and to serve as an irradiation facility for development of fuels and structural materials fur fast reactors. Nuclear reactors pose difficulties to the NDT techniques used to monitor the conditions of the internal components. Sodium cooled fast breeder reactors have their own typical difficulties in using the NDT techniques. These are due to the need for operation in aggressive environment of nuclear radiation and sodium (molten/vapour), as well as the need to maintain leak tightness of a very high order during all states of reactor operation and shutdown for fuel handling, maintenance and remote inspection. This paper discusses the following NDT techniques, which have been successfully used for the past 15 years in FBTR: (i) Periscope and Projector, (ii) Core Co-ordinate Measuring Device and, (iii) Optical fiberscope. The inspection using these techniques have given confidence for further reactor operation at high power by giving useful data on the conditions of the components inside the reactor vessel.

• 대한기계학회논문집A
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• 제34권10호
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• pp.1455-1462
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• 2010

초고온가스로에서 생성된 $950^{\circ}C$ 정도의 초고온 열을 이용하여 수소를 경제적이며 또한 대량으로 생산하기 위한 시스템이 원자력수소생산시스템이며, 이 시스템에서 공정열교환기는 초고온 열과 황-요오드 공정을 통해 수소를 생산하는 핵심 기기이다. 한국원자력연구원에서는 초고온가스로에 사용될 기기에 대한 성능시험을 위해 헬륨가스루프를 구축하고 공정열교환기 시제품을 제작하였다. 본 연구는 공정열교환기 시제품을 헬륨가스루프에서 시험하기 전에 미리 공정열교환기 시제품의 고온 구조건전성을 평가하기 위한 작업의 일환으로 공정열교환기 시제품에 대한 고온구조해석 모델링, 열해석 및 열팽창해석 결과들을 정리한 것이다. 해석 결과는 공정열교환기 시제품 성능시험 장치 설계에 반영할 것이다.

• 한국생물공학회:학술대회논문집
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• 한국생물공학회 2000년도 춘계학술발표대회
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• pp.406-409
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• 2000

본 연구는 T시에서 다량으로 배출되고 있는 종합염색폐수의 생물학적인 처리방법을 기존의 활성슬러지조에 비해 별도의 산기장치가 없는 상태에서도 산소전달효율이 뛰어나며, 기포의 표면적이 작아 용존 되어지는 산소의 양이 많고 또한, 저동력비을 갖는 JLR를 사용하여 새로운 생물학적 처리 공정을 확립하고자 하였다. 이를 이용하여 BOD의 제거효율이 짧은 8시간의 체류시간에서도 99%에 이르는 뛰어난 유기오염원의 제거효율을 얻었다. 산업의 발달로 의식수준이 점차 높아져감에 따라 생활환경이 개선되어 다량의 난분해성 물질이 주요원인인 색도가 수질의 큰 문제점으로 대두되었는데, 이를 제거하는 방법에 있어서도 기존의 1차 응집공정의 과량의 응집제 사용으로 인한 대량의 슬러지생산과 처리비용 등에 관한 문제점을 효율이 뛰어난 활성탄 담체를 포함한 JLR로서 1차 처리함으로써 응집제의 양의 감소와 87%에 달하는 색도의 제거효율을 얻을 수 있었다. 이를 다시 2차 응집하면 약 2,300의 유입수를 150으로 94%정도의 제거효율을 얻어 수질의 향상시킬 수 있었다.