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이재선(Jae-Seon Lee),김종인(Jong-In Kim),김지호(Ji-Ho Kim),박홍윤(Hong-Yune Park),김긍구(Keung-Koo Kim) 한국트라이볼로지학회 2004 한국트라이볼로지학회 학술대회 Vol.38 No.-
Water-lubrication ball bearings are. required to install in aqueous medium where water is used as coolant or working fluid. However water-lubricated frictional characteristics of stainless steel ball bearing is not well known compared to oil-lubricated frictional characteristics. Furthermore study on friction at high temperature is rare because bearing maintenance strategy for water-lubricated or chemicals-lubricated bearings of equipment is mostly based on change of failed bearings and parts. Ball bearings and ball screws are installed on the power transmission for a developing integral reactor and these are lubricated with high temperature and high pressure chemically-controlled water. Bearings and power transmitting mechanical elements for an atomic reactor needs high reliability and high performance during estimated lifetime, and their performance should be verified. In this paper, experimental research results of frictional characteristics of water-lubricated ball bearing are reported.
이재선(Jae-Seon Lee),김종인(Jong-In Kim),김지호(Ji-Ho Kim),박홍윤(Hong-Yune Park),지성균(Sung-Qunn Zee) 한국트라이볼로지학회 2003 한국트라이볼로지학회 학술대회 Vol.37 No.-
Water-lubricated frictional characteristics of stainless steel ball bearing is not well known compared to oil-lubricated frictional characteristics. Furthermore study on friction at high temperature is rare because bearing maintenance strategy for water-lubricated or chemicals-lubricated bearings of equipment is mostly based on change of failed bearings and parts. Ball bearings and ball screw are installed on the power transmission for a developing integral reactor and these are lubricated with high temperature and high pressure chemically-controlled pure water. Bearings and power transmitting mechanical elements for an atomic reactor needs high reliability and high performance during estimated lifetime, and it should be verified. In this paper, experimental research results of frictional characteristics of water-lubricated ball bearing as a preliminary investigation.
이형연(Hyeong-Yeon Lee),어재혁(Jae-Hyuk Eoh),김종범(Jong-Bum Kim),박홍윤(Hong-Yune Park) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.5-2
In a Korean Generation IV demonstration sodium-cooled fast reactor (SFR), various types of high temperature heat exchangers such as IHX (Intermediate heat exchanger), DHX (Decay heat exchanger), AHX (Air heat exchanger) and SG (Steam Generator) are to be designed and installed. In this study, high temperature conceptual design and damage evaluation for the sir-to-sodium heat exchanger, AHX operating at creep regime in a 600MWe Demonstration Sodium-cooled Fast Reactor (SFR) have been performed according to recent version of the high temperature design codes of the ASME Section Ⅲ Subsection NH and RCC-MR. Evaluations of creep-fatigue damage based on full 3D finite element analysis were conducted according to the high temperature design codes and code comparisons were made.
원자력 수소 생산용 공정열교환기 모델의 고온설계 및 손상 평가
이형연(Hyeong-Yeon Lee),송기남(Kee-Nam Song),홍성덕(Sung-Deok Hong),김용완(Yong-Wan Kim),박홍윤(Hong-Yune Park) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.11
The generation of hydrogen using nuclear energy has become a viable means in producing an emission-free energy. A process heat exchanger (PHE) transfers the heat generated from a nuclear hydrogen reactor to a sulfur-iodine hydrogen production system in the NHDD (Nuclear Hydrogen Development and Demonstration). The inlet temperature in the primary side of the PHE mockup in this study is 950℃ with an internal pressure of 7MPa while the inlet temperature in the secondary side of the PHE is 500℃ with internal pressure of 4MPa. An evaluation of creep-fatigue damage on the PHE mockup has been carried out from finite element analysis with the full three dimensional model of the PHE. The candidate materials of the PHE are Alloy 617 and Hastelloy X. In this study, Alloy 617 was considered as PHE material because a very high temperature design guideline is currently available only for Alloy 617 as ASME draft code case. Based on the 3D finite element analysis on the PHE model, creep-fatigue damage evaluation at very high temperature was carried out according to the ASME Draft Code Case for Alloy 617.
소듐냉각 고속로 열교환기의 고온 설계 및 해석 (Ⅰ) - DHX
이형연(Hyeong-Yeon Lee),김종범(Jong-Bum Kim),어재혁(Jae-Hyuk Eoh),이용범(Yong-Bum Lee),박홍윤(Hong-Yune Park) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.4
본 논문에서는 소듐냉각 고속로(sodium-cooled fast reactor, SFR)의 ‘SFR 안전성 모의시험시설’ 개발과 관련하여 현재 한국원자력연구원이 개발 중인 STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) 시험장치의 붕괴열제거 열교환기(DHX)의 고온설계 및 크리프-피로 손상평가에 대해 고찰하였다. DHX, AHX 및 일차열전달계통(PHTS) 기계식 펌프 등의 주기기 성능시험 및 코드 검증을 위한 STELLA 시험루프 내 설치되는 DHX의 고온 설계 및 손상 평가결과 현재의 설계개념은 Mod.9Cr-1Mo 재질의 DHX 기기에 대해 고온 설계 요건을 만족하는 것으로 나타났다. Mod.9Cr-1Mo강의 DHX 본체 노즐과 316SS 재질의 배관이 접합되는 이종용접부에 대해서는 상세 설계 및 고온 구조평가를 향후 수행할 예정이다. The high temperature hoot exchangers in design of Korean Generation Ⅳ demonstration sodium-cooled fast reactor are IHX (Intermediate heat exchanger), DHX(Decay heat exchanger), AHX(Air heat exchanger) and SG(Steam Generator). The tube shapes of the IHX and DHX are straight while those of the SG and AHX are helical. A sodium test facility of the ‘STELLA’(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) is under construction at KAERI for the performance tests and computer code verification of heat exchangers as well as mechanical pump. In this study, high temperature design and evaluation of creep-fatigue damage for the DHX operating at creep regime of 510℃ have been conducted according to the ASME-NH and RCC-MR code. Code comparisons were conducted for the Mod.9Cr-1 Mo steel DHX.