• 한국압력기기공학회 논문집
• /
• 제17권1호
• /
• pp.56-63
• /
• 2021

The feedwater ring is an assembly in steam generator internal piping, which distributes feedwater into the secondary side of the steam generator. It consists of an assembly of carbon steel piping, pipe fittings and J-nozzles which are inserted into the top of the feedwater ring and welded to the diameter of the ring. The feedwater ring at the attachment region of the J-nozzle may be susceptible to flow accelerated corrosion (FAC) due to flow turbulence which increases local fluid velocities. If a J-nozzle becomes a loose part, it can cause damage to tubing near the tube sheet. In this paper, the structural stress analysis for a wall thinned feedwater ring and integrity evaluations under assumed loading conditions are carried out in compliance with ASME B&PV SecIII, NB-3200.

• 대한용접접합학회:학술대회논문집
• /
• 대한용접접합학회 1998년도 특별강연 및 추계학술발표 개요집
• /
• pp.288-291
• /
• 1998

Inconel 600 alloy is used as the material of nuclear steam generator tubing because of its mechanical properties, formability, and corrosion properties. According to reports, the life time of nuclear power plants decreases because of the pitting, intergranular attack, primary water stress corrosion cracking(PWSCC), and intergranular stress corrosion cracking(IGSCC), and denting in the steam generator. The SCC test is very important because of SCC appears in various environment such as solutions, materials, and stress. The C-Rig specimen was made of the steam generator welded sleeve repairing by the pulsed Nd:YAG laser. In the corrosion invironment, corrosion solutions are Primary Water, Caustic, and Sulfate solution and corrosion time is 1624-4877hr. The permitted stress is 30-60ksi.In this C-Ring SCC test is the relationship between corrosion depth, crack and corrosion environment is evaluated. SCC was happens in Sulfate and Corrosion solution but doesn't happen in Primary Water. The corrosion time and stress is very affected by the severely environment of Sulfate or Caustic solution. The microstructure observation indicates that SCC causes interganular failure in the grain boundary of vertical direction.

• 한국압력기기공학회 논문집
• /
• 제15권2호
• /
• pp.64-70
• /
• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2003년도 춘계학술발표대회 요약집
• /
• pp.249-249
• /
• 2003

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.345-346
• /
• 2005

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.869-870
• /
• 2005

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 추계학술발표회
• /
• pp.943-944
• /
• 2005

• Corrosion Science and Technology
• /
• 제10권4호
• /
• pp.118-124
• /
• 2011

Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) has been degraded by various corrosion mechanism during the long-term operation. Especially lead (Pb) is known to be one of the most deleterious species in the secondary system causing outer diameter stress corrosion cracking (ODSCC). Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that a property change of the oxide formed on SG tubing materials by lead addition into a solution is closely related to PbSCC. In the present work, the SCC susceptibility was assessed by using a slow strain rate test (SSRT) in caustic solutions with and without lead for Alloy 600 and Alloy 690 (Ni 60 wt%, Cr 30 wt%, Fe 10 wt%) used as an alternative of Alloy 600 because of outstanding superiority to SCC. The results were discussed in view of the oxide property formed on Alloy 600 and Alloy 690. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDXS).

• 한국압력기기공학회 논문집
• /
• 제5권2호
• /
• pp.1-6
• /
• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).

• 비파괴검사학회지
• /
• 제20권5호
• /
• pp.451-456
• /
• 2000

원전 증기발생기 전열관에 대한 가동중 와전류 검사의 신뢰성을 높이기 위해서는 전열관에서 발생하는 실제와 동일한 부식결함을 제작한 다음 모의과정을 통하여 얻어지는 신호를 해석, 평가하여 장비 및 검사자의 기량을 검증하고 향상시킬 수 있는 기술개발이 이루어져야 한다. 본 논문에서는 가동안전성의 측면, 모의시편의 관점, 인출 전열관의 파괴검사로부터 도출된 관점 그리고 규제기준 및 외국의 사례를 통하여 부식결함을 이용한 증기발생기 전열관에 대한 와전류 검사 신뢰성 향상 연구의 필요성을 고찰하고, 실험실적인 부식결함 제작 모형을 소개하며 그 활용방안을 제시하였다.