• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2003년도 학술대회지
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• pp.193-200
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• 2003

In this study a general Galerkin FE formulation of the incompressible Reynolds equation is derived for lubrication analyses of noncontacting mechanical face seals. Then, the formulation is applied to analyze the flexibly mounted stator-type reactor coolant pump seals of local nuclear power plants, which have deep straight grooves or plane coning on their primary seal ring faces. Their various lubrication performances have been predicted. Results show that the analyzed deep straight groove seal should have a net coning of less than $0.6\;{\mu}m$ to satisfy the leakage limit. And for the same amount of equilibrium opening force the plane coning seal requires to have a 3 times higher dimensionless coning than the deep straight groove seal.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.30-39
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• 2011

URANUS (Ubiquitous, Robust, Accident-forgiving, Non-proliferating, Ultra-lasting and Sustainer) has been developed with 35MWe (100MWth) operating without primary coolant pump, capitalizing on natural circulation capability of lead-bismuth eutectic (LBE) for long-life small and robust power units. To ensure the structural integrity, the large safety margin against Safe Shutdown Earthquake, 0.3g, and furthermore the cost effectiveness for URANUS, three-dimensional seismic base isolation design has been developed. The analytical model has been developed and seismic time history analyses have been carried out. The advantage for using three-dimensional seismic base isolation for URANUS has been discussed.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.96-101
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• 2004

Cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing and valve bodies in light water reactors. These components are subject to thermal embrittlement at the reactor operating temperature. The objective of this study is to summarize the method of estimating ferrite content, Charpy impact energy and J-R curve and to evaluate the thermal embrittlement of the cast austenitic stainless steel piping used in the domestic nuclear power plants. The result of evaluation, two domestic nuclear power plants used CF-8M and CF-8A material has adequate fracture toughness after saturation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2043-2052
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• 2008

The increasing heat generation rates in CPU of notebook computers motivate a research on cooling technologies with low thermal resistance. This paper develops a closed-loop two-phase cooling system using a micropump to circulate a dielectric liquid(PF5060). The cooling system consists of an evaporator containing a boiling enhancement microstructure connected to a condenser with mini fans providing external forced convection. The cooling system is characterized by a parametric study which determines the effects of volume fill ratio of coolant, existence of a boiling enhancement microstructure and pump flow rates on thermal performance of the closed loop. Experimental data shows the optimal parametric values which can dissipate 33.9W with a film heater maintained at $95^{\circ}C$.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.58-67
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• 1997

A numerical study has been peformed to investigate the flow and mixing characteristics of a chemical injection tank in the chemical and volume control system (CVCS) of Yonggwang 5&6 (YGN 5&6). This study was undertaken to provide a basis for modification of the previous design (YGN 3&4) which gave a lot of difficulties in installation and operation of the chemical injection system during the start-up test because it needs a special reciprocating pump with a high actual head. For the tank of length-to-diameter ratios (L/D) of 1,2 and 3, each with and without a baffle inside, calculation results were obtained by solving the unsteady laminar two-dimensional elliptic forms of governing equations for the mass, momentum and species concentration. Finite-difference method was used to obtain discretized equations, and the SIMPLER solution algorithm, which was developed based on the staggered grid control volume, was employed for the calculation procedure. Results showed that the baffle is very effective in enhancing the mixing in the tank and that a baffle should be installed near the tank entrance in order to 110 chemicals into the reactor coolant system (RCS) within the operating time required.

• 대한기계학회논문집A
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• 제28권4호
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• pp.460-466
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• 2004

Cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing and valve bodies in light water reactors. These components are subject to thermal aging at the reactor operating temperature. Thermal aging results in spinodal decomposition of the delta-ferrite leading to increased strength and decreased toughness. This study shows that ferrite content can be predicted by use of the artificial neural network. The neural network has trained teaming data of chemical components and ferrite contents using backpropagation learning process. The predicted results of the ferrite content using trained neural network are in good agreement with experimental ones.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2006년도 하계학술발표대회 논문집
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• pp.471-476
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• 2006

Thermoacoustic refrigerators are operated with acoustic power to pump heat. The acoustic standing wave displaces the gas In the channels of the stack while compressing and expanding. The thermal interaction between the osillating gas he surface of the stack generates an acoustic heat pumping. in this study, a thermoacoustic refrigerator is composed of a resonator of 4cm diameter, stack of plates, heat exchangers and cooling part. Length of the hot heat exchangers, the stack of plates and the cold heat exchanger are 9mm, 8mm and 6mm respectively. Using helium as a coolant at frequency of 516Hz, the cold-part temperature of exchanger fell to $-19.0^{\circ}C$ after 1hours.

• 대한기계학회논문집B
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• 제40권6호
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• pp.403-408
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• 2016

리튬 이온 배터리는 높은 에너지 밀도와 안정적인 충전/방전 특성을 내재하고 있어 하이드리드 및 전기자동차에 보편적으로 사용된다. 리튬 이온 배터리의 효율은 배터리 자체의 온도 특성에 직접적인 영향을 받으므로, 열을 효율적으로 냉각하는 기술이 요구된다. 본 논문에서는 수냉식 배터리 냉각 시스템의 냉각 성능과 펌프 소모동력에 관한 전산유체해석을 수행하였다. 이를 위해 배터리 셀의 냉각수 유량 및 냉각 채널의 특성에 따른 냉각 성능을 수치적으로 예측하였다. 이를 바탕으로 250개 배터리 셀을 기준으로 유량 및 차압에 의한 소모동력을 계산하였다. 이러한 연구는 차세대 하이브리드 및 전기자동차의 시간에 따른 배터리의 온도 변화 및 충/방전 효율 최적화 기술에 적용할 수 있는 기초 연구로 활용될 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Journal of Radiation Protection and Research
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• 제28권2호
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.