• Journal of the Korean Wood Science and Technology
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• 제29권2호
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• pp.69-75
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• 2001

우드세라믹은 목재나 목질재료에 페놀수지를 함침한 후 고온에서 소성하여 만든 새로운 다공질 탄소재료이다. 이러한 우드세라믹의 성질은 재료의 밀도에 영향을 많이 받는 것으로 알려져 있다. 본 연구에서는 증기분사를 하지 않은 보드와 증기분사를 1분, 5분, 10분 실시한 보드 등 4가지 타입의 보드를 만든 다음 우드세라믹을 제조하여 증기분사 시간에 따른 우드세라믹의 성질을 검토하였다. 그 결과 우드세라믹의 중량감소율은 4가지 타입 모두 밀도가 증가함에 따라 증가하는 경향을 보였으며, 중량, 길이 및 두께 감소율은 증기분사를 하지 않은 보드로 제조된 우드세라믹이 증기분사를 실시한 보드로 제조된 우드세라믹보다 약간 크게 나타났다. 또한 증기분사를 하지 않은 보드로 제조된 우드세라믹의 휨강도 및 압축강도가 컸으나 증기분사 시간에 따라서는 뚜렷한 차이가 없는 것으로 나타났다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.695-700
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• 1997

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of .each strategy are also investigated.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1677-1688
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• 2020

In the sodium-cooled fast reactor, the steam generator is a heat exchange device between sodium and water, which may cause leakage, resulting in a sodium-water reaction accident, which in turn affects the safe operation of the entire nuclear reactor. To this end, the electromagnetic vortex flowmeter is used to detect leakage of the steam generator and its signal processing method is studied in this paper. The hydraulic experiment was carried out by using water instead of liquid sodium, and the sensor output signal of the electromagnetic vortex flowmeter under different gas injection volumes was collected. The bubble noise signal is reflected by the base line of the sensor output signal. According to the relationship between the proportion of the bubble noise signal in the sensor output signal and the gas injection volume, a signal processing method based on the energy ratio calculation is proposed to detect whether the water contains bubbles. The gas injection experiment of liquid sodium was conducted to verify the effectiveness of the signal processing method in the detection of bubbles in sodium, and the minimum detectable leak rate of water in the steam generator was detected to be 0.2 g/s.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.727-734
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• 2012

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.95-103
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• 2019

Vapor explosions can be classified in terms of modes of contact between the hot molten fuel and the coolant, since different contact modes may affect fuel-coolant mixing and subsequent vapor explosion energetics. It is generally accepted that most vapor explosion phenomena fall into three different modes of contact; fuel pouring into coolant, coolant injection into fuel and stratified fuel-coolant layers. In this study, we review previous stratified steam explosion experiments as well as recent experiments performed at the KTH in Sweden. While experiments with prototypic reactor materials are minimal, we do note that generally the energetics is limited for the stratified mode of contact. When the fuel mass involved in a steam explosion in a stratified geometry is compared to a pool geometry based on geometrical aspects, one can conclude that there is a very limited set of conditions (when melt jet diameter is small) under which a steam explosion is more energetic in a stratified geometry. However, under these limited conditions the absolute energetic explosion output would still be small because the total fuel mass involved would be limited.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.661-662
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• 2005

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• Structural Engineering and Mechanics
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• 제66권4호
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• pp.505-513
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• 2018

Steam flooding is widely used in heavy oil reservoir with coupling effects among the formation temperature change, fluid flow and solid deformation. The effective stress, porosity and permeability in this process can be affected by the multi-physical coupling of thermal, hydraulic and mechanical processes (THM), resulting in a complex interaction of geomechanical effects and multiphase flow in the porous media. Quantification of the state of deformation and stress in the reservoir is therefore essential for the correct prediction of reservoir efficiency and productivity. This paper presents a coupled fluid flow, thermal and geomechanical model employing a program (MATLAB interface code), which was developed to couple conventional reservoir (ECLIPSE) and geomechanical (ABAQUS) simulators for coupled THM processes in multiphase reservoir modeling. In each simulation cycle, time dependent reservoir pressure and temperature fields obtained from three dimensional compositional reservoir models were transferred into finite element reservoir geomechanical models in ABAQUS as multi-phase flow in deforming reservoirs cannot be performed within ABAQUS and new porosity and permeability are obtained using volumetric strains for the next analysis step. Finally, the proposed approach is illustrated on a complex coupled problem related to steam flooding in an oil reservoir. The reservoir coupled study showed that permeability and porosity increase during the injection scenario and increasing rate around injection wells exceed those of other similar comparable cases. Also, during injection, the uplift occurred very fast just above the injection wells resulting in plastic deformation.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• 청정기술
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• 제29권4호
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• pp.327-332
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• 2023

본 연구에서는 바이오가스 수증기 개질을 통한 합성가스 생산에 관해 연구했다. Ni-Al₂O₃ 및 Ni-CeO₂ 촉매는 공침법으로 제조되었으며 침전제 주입 속도가 조절되었다. 온도에 따른 촉매 성능 테스트는 CH₄:CO₂:H₂O = 1:0.67:3의 가스 조성비와 647,000 mL h^-1 gcat^-1의 공간속도에서 진행하였다. 침전제 주입 속도는 촉매 특성에 영향을 미쳤으며, 지지체 종류에 따라 결과를 보였다. 온도가 증가함에 따라 이산화탄소 개질 반응과 reverse water gas shift 반응이 일어났다. 수증기가 충분히 공급되는 조건에서는 침전제를 한 번에 투입하여 제조된 Ni-Al₂O₃ 촉매가 높은 Ni 분산도에 기인하여 가장 우수한 성능을 보였다.