• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.377-385
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• 1998

Many nuclear power plants have experienced unscheduled shutdown due to the leakage of steam generator tubes. The leakages are normally due to the crack, possibly stress corrosion cracking (SCC) near the tube expansion at the top of tubesheet or at the tangential point of the row-1 U-bend region. The conventional eddy current technique, which makes use of a differential bobbin coil, has been found to be inadequate for the early detection of SCC. During the in-service inspection, therefore, it is a general practice that the rotating pancake coil (RPC) is used for detecting the cracks. Even in using RPC, however, it is difficult to determine the depth of the cracks quantitatively. This paper attempts to determine the detectability and sizing ability of RPC technique for axial or circumferential cracks at the tube expansion region. The simulated cracks with various dimensions were fabricated by electro-discharge machining (EDM) method. Experimental results are discussed with theoretical calculations.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09a
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• pp.587-588
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• 2006

The stress corrosion cracking (SCC) susceptibility of Alloy 600 MA, Alloy 600 TT, Alloy 800, and Alloy 690 TT were investigated in a deaerated 0.01 M solution of sodium tetrathionate using reverse u-bend test samples at $340^{\circ}C$. The results showed that SCC occurred in all alloys, excluding Alloy 690 TT. The SCC susceptibility decreased with an increase in the chromium content of the alloys. The results of the deposits and spectra taken from an energy dispersive X-ray system confirmed the existence of a reduced sulfur causing SCC.

• Journal of the Society of Naval Architects of Korea
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• v.29 no.4
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• pp.227-233
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• 1992

To study the large displacement and large rotation problems, geometrically nonlinear formulation of eccentric degenerated beam element has been developed, where the restrictions of infinitesimal rotation increments are removed and the incremental equations are derived using the Taylor series expansion of the displacement function at time t+dt. The geometrically nonlinear analyses are carried out for the cases of cantilever, square frame, shallow arch and 45-degree bend beam and all of them are compared with each of the other results published. The element developed in the present research can be efficiently utilized for analysis of the nonlinear behaviours of structures when displacements and rotations are large.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• Journal of Power System Engineering
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• v.6 no.3
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• pp.24-30
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• 2002

The deformation and fracture behaviors of both Al2O3 and Ni 4.5wt.%Al plasma thermal spray coating were investigated by an acoustic emission method. Plasma thermal spray coating is formed by a process in which melted particles flying with high speed towards substrate, then crash and spread on the substrate surface cooled and solidified in a very short time, stacking of the particles makes coating. A tensile test is conducted on notch specimens in a stress range below the elastic limit of substrate. A bendind test is done on smooth specimens. The waveforms of AE generated from the both test coating specimens can be classified by FFT analysis into two types which low frequency(type I) and high frequency(type II). The type I waveform is considered to corresponds exfoliation of coating layers and type II waveform corresponds the plastic deformation of notch tip. The fracture of the coating layers can estimate by AE event and amplitude, because AE features increase when the deformation generates.

• Corrosion Science and Technology
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• v.17 no.3
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• pp.116-122
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• 2018

The secondary coolant of a nuclear power plant has small amounts of various impurities (S, Pb, and Cl, etc.) introduced during the initial construction, maintenance, and normal operation. While the concentration of impurities in the feed water is very low, the flow of the cooling water is restricted, so impurities can accumulate on the Top of Tubesheet (TTS). This environment is chemically very complicated and has a very wide range of pH from acidic to alkaline. In this study, the characteristics of the oxide and the mechanism of stress corrosion cracking (SCC) are investigated for Alloy 690 TT in alkaline solution containing Pb, Cl, and S. Reverse U-bend (RUB) specimens were used to evaluate the SCC resistance. The test solution comprises 3m NaCl + 500ppm Pb + 0.31m $Na_2SO_4$ + 0.45m NaOH. Experimental results show that Alloy 690 TT of the crevice environment containing Pb, S, and Cl has significant cracks, indicating that Alloy 690 is vulnerable to stress corrosion cracking under this environment.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.8 s.251
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• pp.993-1000
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• 2006

Operating experience of steam generators has shown that cracks of various morphology frequently occur in the steam generator tubes. These cracked tubes can stay in service if it is proved that the tubes have sufficient safety margin to preclude the risk of burst and leak. Therefore, integrity assessment using exact limit load solutions is very important for safe operation of the steam generators. This paper provides global and local limit load solutions for surface cracks in the steam generator tubes. Such solutions are developed based on three-dimensional (3-D) finite element analyses assuming elastic-perfectly plastic material behavior. For the crack location, both axial and circumferential surface cracks, and for each case, both external and internal cracks are considered. The resulting global and local limit load solutions are given in polynomial forms, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.78-83
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• 1998

Alloy 600과 Alloy 690의 응력부식파괴(Stress corrosion cracking, SCC)에 미치는 TiO$_2$의 영향을 315$^{\circ}C$의 10%NaOH 수용액에서 RUB(reverse U-bend) 시편, C-Ring 시편과 CT(compact tension)시편을 사용하여 평가하였다. 시편은 alloy 600 MA(mill anneal), alloy 600 TT(thermal treatment) 그리고 alloy 690 TT로 제작하였다. SCC 시험은 탈산된 10%NaOH 수용액에 2 g/1 TiO$_2$를 첨가한 용액과 첨가하지 않은 용액에서 수행하였으며, 이 조건에서 분극곡선도 얻었다. SCC 시험시 시편을 부식전위로부터 +150 ㎷ 양극분극을 가하였다. 기준전극으로 external Ag/AgCl electrode를 사용하였다. Alloy 600 MA로 제작한 RUB 시편은 TiO$_2$가 없는 용액에서 5일 안에 벽 관통 균열을 보였으나 TiO$_2$가 첨가된 용액에서는 균열을 관찰할 수 없었다. TiO$_2$가 첨가됨에 따라 alloy 600과 alloy 690의 임계전류밀도는 크게 감소하였고 또한 부동태 전류밀도도 감소하였다. 부동테 영역에서 TiO$_2$가 있는 용액의 경우 여러 peak가 있는 반면에 TiO$_2$가 없는 용액은 peak가 뚜렷하지 않았다. 이런 결과는 TiO$_2$가 첨가점에 따라 active region에서도 안정한 부동태 피막이 존재한다는 것을 시사한다. 또한 TiO$_2$가 없는 경우 SCC가 잘 일어나는 영역에 존재하는 부동태 피막이 TiO$_2$ 첨가에 따라 repassivation kinetics 등의 성질이 변화한 것으로 판단된다.

• Korean Journal of Materials Research
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• v.6 no.9
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• pp.885-899
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• 1996

초강인 AISI 4340강을 85$0^{\circ}C$에서 2시간 동안 오스테나이징 처리 후 수냉하고, 250, 400, $600^{\circ}C$에서 각각 2시간 동안 템퍼링 처리를 하였다. AISI 4340강의 인장 특성은 상온에서 측정되었다. AISI 4340강 위에 니켈 전해도금된 것과 도금되지 않은 시편의 분극 특성이 3.5wt%NaCI 수용액과 인공해수에서 측정되었다. AISI 4340강위에 니켈 전해도금된 시편은 500mV(vs. Ag/AgCI)이하의 전위에서 부식 저항이 크게 향상되었다. 그러나 1A/$\textrm{cm}^2$의 전류밀도에서 30분 이상 니켈 전해도금된 시편은 도금층에 불순물과 기공이 형성되었기 때문에 AISI 4340강의 부식 저항은 감소되었다. AISI 4340강의 수소취화형 응력부식균열을 여러 작용 응력과 음극인가전력에서 U-bend 시편을 이용하여 IN 3.5 wt% NaCI 수용액에서 조사되었고, 수소취화형 응력부식균열 거동은 주사전자현미경으로 조사되었다.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.11a
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• pp.84-91
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• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.