• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.269-273
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• 2015

The study investigated the effect of the distance between two collinear circumferential surface cracks on the primary stress corrosion crack (PWSCC) growth in alloy 600TT steam generator tubes using a finite element damage analysis based on the PWSCC initiation model and macroscopic phenomenological damage mechanics approach. The damage analysis method was verified by comparing the results to the previous study results. The verified method was applied to collinear circumferential surface PWSCCs. As a result, it was found that the collinear cracks showed earlier coalescence and penetration times than the a single crack, and the times increased with the distance. In addition, it is expected that penetration may occur before coalescence of two cracks if they are more than a specific distance apart.

• Transactions of the Korean Society of Automotive Engineers
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• v.10 no.4
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• pp.141-148
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• 2002

Repair were made for front pillar, center pillar and side-step panel for lightweight vehicles with head-on and 40% off-set collision of 15 km/h in a RCAR standard. The salt dilution was sprayed and the compression tests were performed for vehicles with and without anti-corrosional treatment after repair. After 764 hours of salt-dilt sprayed test without using anti-corrosion, the mean penetration depth fur corrosion was shown to be 58% of the thickness. The resulyed decrease in bending stiffness by 10∼20% can cause reduction of the residual life and crash-absorption capability for damaged vehicles. The corrosoin safety tests showed that the anti-corrosional treatment should be made to improve the safety characteristics for a or damaged car.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.3
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• pp.387-395
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• 2013

Recently, several circumferential cracks were found in the control rod drive mechanism (CRDM) nozzles of U.S. nuclear power plants. According to the accident analyses, coolant leaks were caused by primary water stress corrosion cracking (PWSCC). The tensile residual stresses caused by welding, corrosion sensitive materials, and boric acid solution cause PWSCC. Therefore, an exact estimation of the residual stress is important for reliable operation. In this study, finite element simulations were conducted to investigate the effects of the tube geometry (thickness and radius) on the residual stresses in a J-groove weld for different CRDM tube locations. Two different tube locations were considered (center-hole and steepest side hill tube), and the tube radius and thickness variables ($r_o/t$=2, 3, 4) included two different reference values ($r_o$=51.6, t=16.9mm).

• Journal of Korean Society of Steel Construction
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• v.20 no.2
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• pp.313-321
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• 2008

Recently in Japan, fracturing was observed on the diagonal member of a through truss bridge at the boundary region with the concrete slab. Local corrosion damage where the diagonal member was enclosed in the concrete slab is an important factor in the fracture. In this study, accelerated exposure tests were carried out on concrete-steel model specimens simulating steel members at the boundary with concrete. Fatigue tests were then performed on the corroded model specimens. Accelerated exposure tests of the S6-cycle, which is carried out on the model specimens for 150, 300, 450 and 600 da ys. Their surface geometry was then measured. From the accelerated exposure test results, change in maximum and mean corrosion depths was determined according to the testing periods. The effect of local corrosion on fatigue strength was also presented based on the fatigue test results.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.90-95
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• 2004

In this study, we proposed new method to mitigate tensile welding residual stress for preventing PWSCC in CRDM nozzle. Residual stress analysis using finite element method is performed to confirm benefit of the new method. In case of applying existing method, tensile axial residual stress decrease by about 28% and tensile hoop residual stress decrease by about 33%. In case of applying the new method, tensile axial residual stress decrease by about 32% and tensile hoop residual stress decrease by about 43%. Therefore, we conclude the new proposed method is more effective to prevent PWSCC than existing method.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.302-307
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• 1989

Nuclide release through penetrations in radioactive waste container is analyzed. Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant nuclide release, but not with the way these penetrations occur. Numerical illustrations show that mass transport from multiple holes can be significant and may approach the mass transfer rate calculated from bare waste forms. Although partially-failed containers may present an important long-term barrier to release of radionuclides, numerous small holes on a container surface have the potential of bypassing the effectiveness of these barriers.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1159-1168
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• 2013

In pressurized water nuclear reactors (PWRs), the reactor pressure vessel (RPV) upper head contains penetration nozzles that use a control rod drive mechanism (CRDM). The penetration nozzle uses J-groove weld geometry. Recently, the occurrence of cracking in alloy 600 CRDM penetration nozzle has increased. This is attributable to primary water stress corrosion cracking (PWSCC). PWSCC is known to be susceptible to the welding residual stress and operational stress. Generally, the tensile residual stress is the main factor contributing to crack growth. Therefore, this study investigates the effect on weld residual stress through different analysis methods for normal operating conditions using finite element analysis. In addition, this study also considers the effect of repeated normal operating condition cycles on the weld residual stress. Based on the analysis result, this paper presents a normal operating condition analysis method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.34-41
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• 2010

This work is concerned with the analysis of stress corrosion cracking(SCC) behavior of Alloy 600 nozzle penetration mock-up according to a residual stress induced by a dissimilar metal welding(DMW) in a nuclear reactor pressure vessel. The effects of the dimension and materials of the nozzle penetration on the deformation and the residual stress induced by DMW were investigated using a finite element analysis(FEA). The inner diameter(ID) change of the nozzle by DMW and its dependance on the design variables, calculated by FEA, were well consistent with those measured from the mock-up. Accelerated SCC tests were performed for three mock-ups with different wall thicknesses in a highly acidic solution to investigate mainly the effect of the residual stress on the SCC behavior of Alloy 600 nozzle. From a destructive examination of the mock-up after the tests, the SCC behavior of the nozzle was fairly related with the residual stress induced by DMW : axial cracks were found in the ID surface of the nozzle within the J-weld region where the highest tensile hoop stress was predicted by FEA, while circumferential cracks were observed beyond both J-weld root and toe where the highest tensile axial stress was expected.