• Corrosion Science and Technology
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• 제3권6호
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• pp.233-239
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• 2004

Primary water stress corrosion cracking of steam generator tubings occur on many tubes in pressurized water reactors(PWRs), and they are repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to know the leak behavior of the tubes, which have stress corrosion cracks. Crack development tests were carried out on the tubes at room temperature, and leak rate and burst pressure were measured on the degraded tubes at room temperature and a high temperature. No leakage was detected on the tubes where 100 % through wall crack developed, at 1560 psi, which is an operating pressure difference of pressurized water reactors(PWRs). In some tests, leak rates of the tubes increased with time at a constant internal water pressure. A test tube showed a very small amount of leakage at 2700 psi in a high temperature pressure test at $282^{\circ}C$, but it disappeared after the pressure increased slightly. Even cracks are 100 % through wall, they need to open in order to reach a certain amount of leak rate at the operating pressure difference.

• Journal of Welding and Joining
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• 제9권4호
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• pp.60-68
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• 1991

The weld quality of end cap welds in Pressurized Heavy Water Reactor (PHWR) Fuel is extremely important for the fuel performance in the nuclear reactor. The quality of resistance upset welds is currently evaluated mainly by the metallographic examination although it reveals only two weld cross-sections in a circumference welds. This investigation was, firstly, carried out to determine whether the ultrasonic examination would be applied to detect weld defects in the end cap welds and, secondly, to measure the mechanical strength of upset butt welds as a function of phase shift percentage. The major results obtained in this study are as follows: 1. The weld current and amount of upset shrinkage linearly increased with increasing the phase shift percentage. 2. Above the phase shift 55%, the defects in the welds were completely eliminated with increasing the phase of sound weld was over the thickness of cladding tube. 3. The ultrasonic testing well detected such defects in the end cap welds as upset external crack, upset split, corner crack and irregular weld flash comparing with the results of metallography. 4. The micro-fissure in the corner of the end cap welds was reliably detected by ultrasonic testing. 5. The mechanical strength in the welds increased with increasing phase shift percentage but the fracture did't occur in the welds above 55%.

• 대한기계학회논문집A
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• 제26권11호
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• pp.2390-2398
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• 2002

For integrity analysis of nuclear reactor pressure vessel, including the Pressurized thermal shock analysis, the fast and accurate calculation of the stress intensity factor at the crack tip is needed. For this, a simple approximation scheme is developed and the resulting stress intensity factors for axial semi-elliptical cracks in cylindrical vessel under various loading conditions are compared with those of the finite element method and other approximation methods, such as Raju-Newman's equation and ASME Sec. Xl approach. For these, three-dimensional finite-element analyses are performed to obtain the stress intensity factors for various surface cracks with t/R = 0.1. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite clement analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. The approximation solutions are within $\pm$2.5% of the those of FEA using symmetric model of one-forth of a vessel under pressure loading, and 1-3% higher under pressurized thermal shock condition. The analysis results confirm that the VINTIN method provides sufficiently accurate stress intensity factor values for axial semi-elliptical flaws on the surface of the reactor pressure vessel.

• Journal of Welding and Joining
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• 제23권6호
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• pp.61-66
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• 2005

A probabilistic approach for evaluating failure risk is suggested in this paper. Probabilistic fracture analyses were performed for a pressurized pipe of a Cr-Mo steel reflecting variation of material properties at high temperature. A crack was assumed to be located along the weld fusion line. Probability density functions of major variables were determined by statistical analyses of material creep and creep crack growth data measured by the previous experimental studies by authors. Distributions of these variables were implemented in Monte Carlo simulation of this study. As a fracture parameter for characterizing growth of a fusion line crack between two materials with different creep properties, $C_t$ normalized with $C^*$ was employed. And the elapsed time was also normalized with tT, Resultingly, failure probability as a function of operating time was evaluated fur various cases. Conventional deterministic life assessment result was turned out to be conservative compared with that of probabilistic result. Sensitivity analysis for each input variable was conducted to understand the most influencing variable to the analysis results. Internal pressure, creep crack growth coefficient and creep coefficient were more sensitive to failure probability than other variables.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.479-489
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• 2019

Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer. Electrochemical analysis revealed that the oxide layers formed on the LCF crack surface of Alloy 690 had higher impedance and less defect density than those of 316 SS, which resulted in longer LCF life of Alloy 690 than 316 SS in a simulated PWR environment.

• 대한기계학회논문집A
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• 제26권8호
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• pp.1672-1679
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• 2002

This paper provides an engineering J estimation equation for cylinders with finite internal axial surfacecracks under internal pressure. The proposed equation is the J estimation equation based on deformation plasticity using Ramberg-Osgood (R-O) materials. Based on detailed 3-D FE results using deformation plasticity, plastic influence functions for fully plastic J components are tabulated for practically interesting ranges of the mean radius-to-thickness ratio, the crack depth-to-length ratio, the crack depth-to-thickness ratio. the strain hardening index for the R-O material, and the location along the semi-elliptical crack front. Based on tabilated plastic influence functions, the J estimation equation along the crack front is proposed and validated for R-O materials. Good agreements between the FE results and the proposed J estimation provide confidence in the use of the proposed method to elastic-plastic fracture mechanics of pressurized piping.

• 대한기계학회논문집A
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• 제25권4호
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• pp.756-763
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• 2001

In the process of integrity evaluation for nuclear power plant components, a series of fracture mechanics evaluation on surface cracks in reactor pressure vessel(RPV) must be conducted. These fracture mechanics evaluation are based on stress intensity factor, K. However, under pressurized thermal shock(PTS) conditions, the combination of thermal and mechanical stress by steep temperature gradient and internal pressure causes considerably high tensile stress at the inside of RPV wall. Besides, the internal pressure during the normal operation produces high tensile stress at the RPV wall. As a result, cracks on inner surface of RPVs may experience elastic-plastic behavior which can be explained with J-integral. In such a case, however, J-integral may possibly lose its validity due to constraint effect. In this paper, in order to verify the suitability of J-integral, tow dimensional finite element analyses were applied for various surface cracks. A total of 18 crack geometries were analyzed, and $\Omega$ stresses were obtained by comparing resulting HRR stress distribution with corresponding actual stress distributions. In conclusion, HRR stress fields were found to overestimate the actual crack-tip stress field due to constraint effect.

• 한국세라믹학회지
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• 제29권12호
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• pp.970-980
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• 1992

${\gamma}-alumina$ membranes were prepared by sol-gel dip coating or pressurized coating of boehmite sols on slipcasted ${\gamma}-alumina$ support tubes. The particle size of sols synthesized via the modified Yoldas-method could be controlled below 5 mn according to the mole ratio of nitric acid/aluminumtri-sec-butoxide (0.07~1.0). The reproducible crack-free composite membranes were produced by the 2nd dip coating or the pressurized coating technique using very stable sols with the particle size of 45 nm. Nitrogen gas permeability through the top-layer in the composite membrane was about $70~55{\times}10^{-7}\;mol/m^2{\cdot}s{\cdot}Pa$. The thermal stability of the top layer was proved to be good enough upto the heat-treatment temperature of $500^{\circ}C$.

• Journal of Mechanical Science and Technology
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• 제20권3호
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• pp.319-328
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• 2006

A simple approximation method for the stress intensity factor at the tip of the axial semielliptical cracks on the cylindrical vessel is developed. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite element analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. For these, 3-D finite-element analyses are performed to obtain the stress intensity factors for various surface cracks with t/R=0.1. The approximation solutions are within $\pm2.5%$ of the those of finite element analysis using symmetric model of one-forth of a vessel under pressure loading, and 1-3% higher under pressurized thermal shock condition. The analysis results confirm that the approximation method provides sufficiently accurate stress intensity factor values for the axial semi-elliptical flaws on the surface of the reactor pressure vessel.

• 전산구조공학
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• 제9권4호
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• pp.189-198
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• 1996

콘크리트 중력댐 상부면의 균열에 작용하는 수압의 영향을 주로 고려하여 댐의 파괴거동을 조사하였다. 첫째, 표면적분법에 의하여 응력확대계수를 구하는 경우에 작용하는 수압의 형태를 등분포형태 외에, 삼각형 분포 및 포물선분포도 고려하여 보았다. 둘째, FRANC(FRacture Analysis Code)를 이용하여 균열면에 작용하는 수압의 형태에 따른 기존균열의 전파방향을 추적하였다. 셋째, 월류수위 아래에서 균열이 전파되지 않을 수 있는 한계균열길이를 수압의 분포형태에 따라 구분하여 구하여 보았다. 표면적분법으로 수압의 형태에 따라 응력확대계수를 구한 결과는 FRANC를 이용하여 얻어진 결과와 비교 되었으며 비교적 잘 일치함을 알 수 있었다. 균열면에 작용하는 수압의 형태가 삼각형분포의 경우에 균열의 전파방향은 등분포의 경우에 비하여 댐의 기초쪽으로 기우는 것을 알 수 있었으며, 또한 월류수위 아래에서 한계균열길이는 댐높이의 대략 2/5-1/2되는 곳에서 최대가 됨을 알 수 있었다.