• 한국기계가공학회지
• /
• 제17권4호
• /
• pp.150-159
• /
• 2018

The present research deals with a finite element analysis and fatigue evaluation of a steam separator of a high-pressure evaporator for the Heat Recovery Steam Generator (HRSG). The fatigue during the expected life of the HRSG was evaluated according to the ASME Boiler and Pressure Vessel Code Section VIII Division 2 (ASME Code). First, based on the eight transient operating conditions prescribed for the HRSG, temperature distribution of the steam separator was analyzed by a transient thermal analysis. Results of the thermal analysis were used as a thermal load for the structural analysis and used to determine the mean cycle temperature. Next, a structural analysis for the transient conditions was carried out with the thermal load, steam pressure, and nozzle load. The maximum stress location was found to be the riser nozzle bore, and hence fatigue was evaluated at that location, as per ASME Code. As a result, the cumulative usage factor was calculated as 0.00072 (much less than 1). In conclusion, the steam separator was found to be safe from fatigue failure during the expected life.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2007년도 춘계학술대회A
• /
• pp.250-255
• /
• 2007

Steam generator tubes play an important role in safety because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. For this reason, the integrity of the tubes is essential in minimizing the leakage possibility of radioactive water. The integrity of the tubes is evaluated based on NDE (non-destructive evaluation) inspection results. Especially ECT (eddy current test) method is usually used for detecting the flaws in steam generator tubes. However, detection capacity of the NDE is not perfect and all of the "real flaws" which actually existing in steam generator tunes is not known by NDE results. Therefore reliability of NDE system is one of the essential parts in assessing the integrity of steam generators. In this study POD (probability of detection) of ECT system for ODSCC in steam generator tubes is evaluated using multivariate logistic regression. The cracked tube specimens are made using the withdrawn steam generator tubes. Therefore the cracks are not artificial but real. Using the multivariate logistic regression method, continuous POD surfaces are evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive evaluation of the cracked tubes. Length and depth of cracks are considered in multivariate logistic regression and their effects on detection capacity are evaluated.

• Nuclear Engineering and Technology
• /
• 제47권3호
• /
• pp.260-266
• /
• 2015

Twenty sodium-cooled fast reactors (SFRs) have provided valuable experience in design, licensing, and operation. This paper summarizes the important safety criteria and safety guidelines of intermediate sodium systems, steam generators, decay heat removal systems and associated construction materials and in-service inspection. The safety criteria and guidelines provide a sufficient framework for design and licensing, in particular by new entrants in SFRs.

• 한국에너지공학회:학술대회논문집
• /
• 한국에너지공학회 1999년도 추계 학술발표회 논문집
• /
• pp.125-130
• /
• 1999

Corrosion of steam generator tubes is the major issue affecting selection of secondary water chemistry parameters. The objective of secondary side water chemistry control is to minimize corrosion damage and to thereby maximize the reliability and economic performance of the secondary system. To achieve this objective, the water chemistry has to be compatible with all parts of the system including steam generators.(Omitted)

• 대한전자공학회:학술대회논문집
• /
• 대한전자공학회 1998년도 하계종합학술대회논문집
• /
• pp.223-226
• /
• 1998

The control objective of steam generator water level in the secondary circuit of a nuclear power plant is to regulate the water level at the desired set point. The dynamics of steam generators is non-linear in nature. The task of modelling such plant is very difficult and especially so when plant operating conditions change frequently. In these reasons, conventional PI gains over all pover range will not work efficiently and a manual control is generally used in low power operation. Therefore the robust H$\infty$ controller design method should be required. In this paper, we design the robust H$\infty$ controller for water level control of steam generator using a mixed H$\infty$ optimization with model-matching method. Firstly we choose the desired model that has good disturbance rejection performance. Secondly we design a stabilizing controller to keep the model-matching error small and also provide sufficiently large stability margin against additive perturbations of the nominal plant.

• 대한기계학회논문집A
• /
• 제28권7호
• /
• pp.1006-1013
• /
• 2004

Operating experience of steam generators shows that the tubes are degraded by stress corrosion cracking, fretting wear and so on. These defected tubes could stay in service if it is proved that the tubes have sufficient structural margin to preclude the risk of tube bursting. This paper provides detailed plastic limit pressure solutions for through-wall cracks in the steam generator tubes. These are developed based on three dimensional(3D) finite element analyses assuming elastic-perfectly plastic material behavior. Both axial and circumferential through-wall cracks in free span and in u-bend regions are considered. The resulting limit pressure solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• 한국압력기기공학회 논문집
• /
• 제4권2호
• /
• pp.29-33
• /
• 2008

As the measures of nuclear power plant utilities and manufacturers to reduce the defects of tube expansion region during manufacturing steam generators, many types of NDEs(Non-Destructive Examinations) are conducted to inspect the expansion region. The expansion region of tube is subject to degrade because of stress concentration induced by tube expansion, sludge pile and high temperature. So the inspections for tube expansion region have been reinforced. Liquid penetrant test, helium leak test, Bobbin profile test and hydraulic test are performed to confirm the integrity of tube expanded by hydraulic expansion method. Liquid penetrant test and helium leak test are used to inspect seal weld region on tubesheet end part. Bobbin Profile test is used to inspect fully the expanded region of steam generator tube. Hydraulic test finally verifies the integrity of seal weld region on tubesheet end part.

• 한국압력기기공학회 논문집
• /
• 제6권2호
• /
• pp.1-8
• /
• 2010

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.

• 한국레이저가공학회지
• /
• 제3권1호
• /
• pp.12-20
• /
• 2000

The nuclear steam generators are subjected to corrosion environmental condition during operation that can result in stress corrosion in the tube wall. If any tube wall degradation is recognized, the tube must be repaired by plugging or sleeving. For the sleeving repair, Nd : YAG laser welded sleeving technology is one of the most promising when considering radioactive working conditions in the nuclear power plant. In this paper, the laser welding characteristics of steam generator tube and sleeve materials are investigated. The effects of average laser power, laser energy, welding speed, pulse duration and frequency are evaluated. Based on these results, Nd : YAG laser welded sleeving repair was applied to the degraded steam generator tubes in real environment.

• Nuclear Engineering and Technology
• /
• 제47권7호
• /
• pp.934-938
• /
• 2015

This technical note provides some considerations for the metallographic observation of intergranular attack (IGA) in Alloy 600 steam generator tubes. The IGA region was crazed along the grain boundaries through a deformation by an applied stress. The direction and extent of the crazing depended on those of the applied stress. It was found that an IGA defect can be misevaluated as a stress corrosion crack. Therefore, special caution should be taken during the destructive examination of the pulled-out tubes from operating steam generators.