• Proceedings of the Membrane Society of Korea Conference
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• 1997.04b
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• pp.50-51
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• 1997

가압경수로형(PWR) 원자력발전소에스는 원자로 출력조절을 위한 중성자 흡수체로 붕산(Boric Acid)을 사용하며, 불순물이 농축되는 것을 방지하기 위하여 이온교환수지로 수질 정화를 하고 있다. 그러나, 붕산으로 포화운전되는 이온교환수지에서 붕산보다 이온선택도가 낮은 실리카는 제거되지 않으므로, 원자력발전소의 운전년수 경과에 따라 1차계통수(원자로냉각재)의 붕산수중에 실리카 농도가 증가하게 된다. 한편, 실리카는 고온, 고압 운전조건에서 양이온불순물과 결합하여 핵연료피복재에 열전달을 방해하는 규석(Zeolite)층을 형성함으로서 국부가열(Hot Spot)에 의한 핵연료 손상을 일으킬 수 있으므로, 효율적인 실리카 제거기술이 요구된다. 따라서, 기존에 원전에서 사용하고 있는 Feed & Bleed에 의한 수질정화 방법은 다량의 폐기물 발생 및 붕산보충이 필요하므로, 역삼투막(RO)을 이용하여 붕소와 실리카의 최적 분리, 회수조건을 연구하고, 붕산저장 용량이 큰 SFP(Spent Fuel Pool)의 수질정화용 이동형 RO장치를 개발하기 위하여 붕산수중의 실리카에 대한 역삼투막의 선택적 제거특성을 검토하였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.25-33
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• 2016

The thermal stratification phenomenon in the nuclear power plant can cause abnormal deformation of the piping, contact with the support, damage to the support system. Repetition of the thermal stratification phenomenon or variation of the thermal boundary layer can cause thermal fatigue. Thermal stratification phenomenon in nuclear power plants is still an ongoing issue and active research has been carried out. In this paper, the current situation in Korean nuclear power plants is described, followed by the status of research and the future problems on the thermal stratification phenomenon in Korea.

• Transactions of the Korean Society of Mechanical Engineers
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• v.13 no.4
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• pp.717-725
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• 1989

An experimental investigation on the combustion behavior of pulverized coal particles was performed using the cross-section micrography techniques while sample coal particles were collected in-situ from the flow reactor. The coal particles were representative of pulverized bituminous coal undergoing a raped pyrolysis and combustion, however, quenched at the time when the particles were deposited onto a sample plate. The internal structure of coal was observed to change as deposited. Upon injection into a flow reactor, bituminous coal particles showed many holes which represented internal pore formation during the pyrolysis. The relative portion of the remaining matrix of coal was decreasing as the residence time progressed. This direct observation of cross-section of burning particles enabled better understanding of the coal combustion behavior.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.1
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• pp.11-17
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• 2015

High-density polyethylene (HDPE) piping, which has recently been applied to safety class III piping in nuclear power plants, can be butt-joined through the thermal fusion process, which heats two fused surfaces and then subject to axial pressure. The thermal fusion process generates bead shapes on the butt fusion. The stress concentrations caused by the bead shapes may reduce the fatigue lifetime. Thus, investigating the effect of the thermal butt fusion beads on fatigue behavior is necessary. This study examined the fatigue behavior of thermal butt fusion via a tensile fatigue test under stress-controlled conditions using finite element elastic stress analysis. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low-cycle fatigue region while having a negligible effect in the medium- and high-cycle fatigue regions.

• Plant Journal
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• v.10 no.2
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• pp.28-37
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• 2014

Recently, it is considered that environment and energy are critical issues all over the world. In power generation sector in Korea, almost power stations are constructed and operated as cogeneration plants in conformity with this trend. KDHC(Korea District Heating Corporation) goes one step further adopting renewable energy technology like heat pump using wasted heat for energy-saving and environment improvement. This study investigates the performance characteristics by the location of waste heat recovery heat pumps of 5 Gcal/h capacity in 150 MW-class Gwang-gyo cogeneration plant using commercial software 'THERMOFLEX'. Prior to analysis, the simulations are performed with actual operation data, and then the validation of simulations is verified by checking the error within 2%. After verification, the simulations are carried out with 3 locations and the effect on electrical power output and heat output is analyzed. As a result, overall efficiency of cogeneration plant is the highest in the case of heat pump located before DH(District Heating) Heater because of the largest increase of heat output despite of decrease of electrical power output.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.4
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• pp.339-345
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• 2008

The heat exchanger tube in nuclear power plants is mainly fabricated from nonferromagnetic material such as a copper, titanium, and inconel alloy, but the moisture separator & reheater tube in the turbine system is fabricated from ferromagnetic material such as a carbon steel or ferrite stainless steel which has a good mechanical properties in harsh environments of high pressure and temperature. Especially, the moisture separator & reheater tubes, which use steam as a heat transfer media, typically employ a tubing with integral fins to furnish higher heat transfer rates. The ferromagnetic tube typically shows superior properties in high pressure and temperature environments than a nonferromagnetic material, but can make a trouble during the normal operation of power plants because the ferrous tube has service-induced damage forms including a steam cutting, erosion, mechanical wear, stress corrosion cracking, etc. Therefore, nondestructive examination is periodically performed to evaluate the tube integrity. Now, the remote field testing(RFT) technique is one of the solution for examination of ferromagnetic tube because the conventional eddy current technique typically can not be applied to ferromagnetic tube such as a ferrite stainless steel due to the high electrical permeability of ferrous tube. In this study, we have designed RFT probes, calibration standards, artificial flaw specimen, and probe pusher-puller necessary for field application, and have successfully carry out RFT examination of the moisture separator & reheater tube of nuclear power plants.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.08a
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• pp.1-1
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• 2010

핵융합에너지는 1930년대 한스 베테에 의해 태양과 별 에너지의 근원임이 밝혀진 후 소핵 폭탄실험 성공으로 그 위력적인 에너지를 인공적으로 만들 수 있음을 세상에 드러내게 된다. 그 뒤 이 에너지의 평화적인 이용 노력이 시작되었고 1958년 스위스에서 핵융합에너지의 평화적 이용에 대한 첫 국제회의가 열리게 되면서 에너지원으로서의 연구를 통해 냉전시대의 경쟁 대상의 과학기술의 하나로 부각되면서 눈부신 성능 향상을 보여주게 되었다. 아직 여러 어려운 관문이 남아있지만 기후변화와 에너지원 고갈에 의한 새로운 에너지원에 대한 강력한 필요성이 제기되면서 ITER와 같은 대형 국제공동연구시설 건설이 시작되었고 2030년대에는 최초의 핵융합발전소를 건설하려는 꿈도 그려가고 있다. 핵융합에너지를 얻는 방식에는 여러 방법이 시도되었는데 현재는 자기장을 이용해 플라즈마를 핵융합반응이 일어나기에 충분한 시간동안 가두는 자기핵융합방식과 관성으로 플라즈마를 가두는 관성핵융합방식으로 크게 구분할 수 있다. 자기핵융합방식의 경우 플라즈마를 만들고 가열하여 핵융합반응 확률이 높은 고온으로 가열하고 그 조건을 오래 지속시키는 기술들이 필요한데 이 기술들은 오늘날의 거의 모든 극한기술들이 망라되어 적용되는데 초전도, 고주파/ 초고주파, 대전력 공급, 대형 시설 실시간 제어기술, 대규모 신호처리기술, 고온 플라즈마 진단 기술, 대규모 시스템 시뮬레이션 기술 등이 그것이다. 여기에 또한 중요한 기술의 하나로 초고진공 기술이 필요하다. 이러한 기술이 집약되고 서로 통합되어 하나의 목적을 위해 쓰여지도록 고안되고 만들어진 장치가 자기핵융합 장치이며 따라서 현대의 자기핵융합장치들은 굉장히 복잡하며 대형 시설로 지어질 수밖에 없다. 우리나라는 1970년대 말부터 소형의 플라즈마 연구시설을 시작으로 자기핵융합 연구를 시작하면서 인력 양성을 시작하였으며 가속기 등 대형 연구시설이 본격적으로 지어지던 1990년대에 세계적으로 유래가 없는 초전도 자기핵융합장치인 KSTAR장치 건설 프로젝트를 시작하게 되었다. 총 11년이 넘는 건설기간 동안 여러 학교와 연구기관, 그리고 산업체가 참여하여 성공적으로 시운전을 실시하였으며 당당히 세계적인 장치를 통한 핵융합연구 대열에 동참하게 되었다. 이를 통한 기술 개발의 결과로 국제적 공동연구장치 ITER의 건설사업에 참여하게 되었고 KSTAR와 ITER를 통해 핵융합 에너지 상용화 기술 개발을 국가적인 기술개발의 목표로 결정하고 연구개발계획을 전략적으로 세워 진행하고 있다. 이번 논문에서는 자기핵융합의 특징과 연구 동향을 통해 우리나라의 기술 수준을 조망하고 특히 진공 기술 분야와의 상호 의존적 영향 분석을 통해 공동의 발전 방향을 모색해 보려고 한다.

• Korean Journal of Soil Science and Fertilizer
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• v.31 no.1
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• pp.61-66
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• 1998

An artificial zeolite was synthesized from bituminous coal fly ash by thermo-alkali treatment, $100{\pm}3^{\circ}C$ and 3.5 N NaOH solution for 24 hours, to develope new useful material in waste water treatment system. The artificial zeolite had much higher cation exchange capacity, $299cmol^+\;kg^{-1}$ and lower concentration of extractable heavy metals than bituminous coal fly ash. The higher the temperature and the longer the shaking time, the more Zn, Cu, Cd and Pb were sorbed chemically by the artificial zeolite. Shaking artificial zeolite and wastewater at $35^{\circ}C$ for 9 hours, the amount of sorbed heavy metals were $123.5g\;kg^{-1}$ for Zn, $164.7g\;kg^{-1}$ for Cu, $184.4g\;kg^{-1}$ for Cd and $350.6g\;kg^{-1}$ for Pb, respectively.

• Journal of Energy Engineering
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• v.28 no.3
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• pp.1-9
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• 2019

A major stress determining the remaining life of the tube in feedwater heater of fossil fuel power plant is hoop stress by the internal pressure. However, thermal stress due to temperature difference across the wall thickness also contributed to reduce the remaining life of the tube. Therefore, thermal loading must be considered even though the contribution of internal pressure loading to the stresses of the tube was known to be much higher than that of the thermal loading. In this study, thermal stress of the tubes in the de-superheating zone was estimated, which was generated due to the temperature difference across the tube thickness. Analytic equations were shown for determining the hoop stress and the radial stress of the tube with uniform thinning and for the temperature across the tube thickness. Accuracy and effectiveness of the analytic equations for the stresses were verified by comparing the results obtained by the analytic equations with those obtained from finite element analysis. Using finite element analysis, the stresses for eccentric thinning were also determined. The effect of heat transfer coefficient on thermal stress was investigated using series of finite element analyses with various values of heat transfer coefficient for both inner and outer surface of the tube. It was shown that the effect of heat transfer coefficient at outer surface was larger than that of heat transfer coefficient at inner surface on the thermal stress of the tube. Also, the hoop stress was larger than the radial stress for both cases of uniformly and eccentrically thinned tubes when the thermal loading was only considered without internal pressure loading.

• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.