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3차원 면진장치를 이용한 URANUS 액체금속로의 지진응답감소
이국희,김윤재,류강묵,황일순,유봉,Lee, Kuk-Hee,Kim, Yun-Jae,Ryu, Kang-Mook,Hwang, Il Soon,Yoo, Bong 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.3
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.
증기발생기 축방향 부분관통균열 전열관의 파열 압력 시험
이국희,김홍덕,강용석,남민우,조남철,Lee, Kuk-Hee,Kim, Hong-Deok,Kang, Yong-Seok,Nam, Min-Woo,Cho, Nam-Cheoul 한국압력기기공학회 2014 한국압력기기공학회 논문집 Vol.10 No.1
In this research, burst tests for axial notched steam generator tubes were conducted. To measure the burst pressure of notched tubes, a burst testing system was manufactured. The tests were conducted under internal pressure at room temperature. Part-through-wall and through-wall notches which have various geometries with different depths and lengths were machined by electro-discharged-machined(EDM) method. The burst pressure decreased exponentially with increasing notch length and decreased almost linearly with increasing notch depth. A comparison of the burst pressure with existing burst pressure solutions for cracked tube show that the existing solution agree well with the test results.
CANDU형 원전 압력관에 존재하는 축방향 균열의 응력확대계수
이국희,오영진,박흥배,정한섭,정하주,김윤재,Lee, Kuk-Hee,Oh, Young-Jin,Park, Heung-Bae,Chung, Han-Sub,Chung, Ha-Joo,Kim, Yun-Jae 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.1
CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.
이국희(Kuk-Hee Lee),최성남(Sung-Nam Choi),김훈태(Hun-Tae Kim) 대한기계학회 2018 대한기계학회 춘추학술대회 Vol.2018 No.12
ASME Boiler and pressure Vessel Code (BPVC) Section XI, Appendix H provides a failure assessment diagram (FAD) procedure for the assessment of ferritic and austenitic piping containing flaws. The evaluation methodology is based on a FAD approach that includes consideration of the three failure mechanisms (brittle fracture, elastic-plastic fracture mechanics and limit load failure). This study provides validation of the ASME BPVC Section XI, Appendix H FAD methodology, via elastic-plastic finite element analyses for axial cracked pipes under internal pressure. It is found that the ASME BPVC FAD agree well with the finite element results.