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한국표준형 원전 증기발생기 전열관 축방향 ODSCC 발생원인 분석
김홍덕,박수기,임창재,정한섭,Kim, Hong-deok,Park, Su-ki,Yim, Chang Jae,Chung, Han Sub 한국압력기기공학회 2010 한국압력기기공학회 논문집 Vol.6 No.1
Domestic nuclear steam generators with Alloy 600 HTMA tubes have experienced axial cracking at eggcrate tube support plates(TSPs). The axial stress corrosion cracks were observed at the crevice between outside of tubes and eggcrate TSPs. The root cause of axial cracking was investigated by thermal hydraulic analysis and sludge distribution diagnosis. It is suggested that deposition of sludge at eggcrate TSPs could increase the outside surface temperature of tube and promote the enrichment of impurities at crevice, and thus accelerate cracking. Additionally strategy for reducing the sludge ingress to steam generators is discussed.
김홍덕,Kim, Hong-deok 한국압력기기공학회 2010 한국압력기기공학회 논문집 Vol.6 No.2
Nuclear steam generator tubes have experienced wear degradation at tube support structure. Morphology of wear scar was analyzed by using eddy current signal. A burst test facility for steam generator tubes was established and tubes with 3 types of defects were tested. The burst test results show that the depth of wear scar is the main factor influencing the burst pressure of tubes, meanwhile, both the longitudinal length and the angle also have effect on the burst pressure. Based on test results, the burst pressure equation for wear degradation was proposed.
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.
이성호,이요섭,김홍덕,이경수,황경모,Lee, Sung Ho,Lee, Yo Seob,Kim, Hong Deok,Lee, Kyoung Soo,Hwang, Kyeong Mo 한국압력기기공학회 2015 한국압력기기공학회 논문집 Vol.11 No.2
Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion, cavitation, flashing and/or liquid droplet impingement, is a main concern in secondary steam cycle piping system of nuclear power plants in terms of safety and operability. Thinned pipe management program (TPMP) has being developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning. In this paper, newest technologies, standards and regulations related to the integrity assessment, repair and replacement of thinned pipe component are reviewed. And technical improvement items in TPMP to secure the reliability and effectiveness are also presented.