• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1412-1422
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• 2016

Austenitic stainless steels (ASSs) are widely used for nuclear pipes as they exhibit a good combination of mechanical properties and corrosion resistance. However, high tensile residual stresses may occur in ASS welds because postweld heat treatment is not generally conducted in order to avoid sensitization, which causes a stress corrosion crack. In this study, round robin analyses on stress intensity factors (SIFs) were carried out to examine the appropriateness of structural integrity assessment methods for ASS pipe welds with two types of circumferential cracks. Typical stress profiles were generated from finite element analyses by considering residual stresses and normal operating conditions. Then, SIFs of cracked ASS pipes were determined by analytical equations represented in fitness-for-service assessment codes as well as reference finite element analyses. The discrepancies of estimated SIFs among round robin participants were confirmed due to different assessment procedures and relevant considerations, as well as the mistakes of participants. The effects of uncertainty factors on SIFs were deducted from sensitivity analyses and, based on the similarity and conservatism compared with detailed finite element analysis results, the R6 code, taking into account the applied internal pressure and combination of stress components, was recommended as the optimum procedure for SIF estimation.

• 비파괴검사학회지
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• 제23권4호
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• pp.364-371
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• 2003

In this paper, a review of the current status, on the use of long range cylindrically guided wave modes, and their interaction with cracks and corrosion damage in pipe-like structures will be discussed. Applications of cylindrically guided ultrasonic wave modes have been developed for inspection of corrosion damage in pipelines at chemical plants, flow-accelerated corrosion damage (wall thinning) in feedwater piping, and circumferential stress corrosion cracks in PWR steam generator tubes. It has been demonstrated that this inspection technique can be employed on a variety of piping geometries (diameters from 1 in. to 3 ft, and wall thickness from 0.1 to 6 in.) and a propagation distance of 100 meters or more is sometimes feasible. This technique can also be used in the inspection of inaccessible or buried regions of pipes and tubes.

• Journal of Welding and Joining
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• 제11권2호
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• pp.42-52
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• 1993

The corrosion fatigue fracture behaviour of dual phase steel was investigated in 3% NaCl solution at 302MPa and 137MPa. Fatigue test was conducted by cantilever type of self-made rotary bending fatigue testing machine. The fatigue strength increased with increasing the hardness of 2nd phase. Corrosion pit originated at the boundary of the 2nd phase. The size and number of corrosion pits were influenced by the 2nd phase hardness, and pits remained constant in size just after they were transited into cracks. The life of crack initiation was effected by stress level. The shape of relation of .DELTA. K and da/dN has smaller scattering in it in 3% NaCl solution than that in air. The higher the 2nd phase hardness is, the greater the corrosion fatigue life becomes. Corrosion fatigue fracture behaviour was primarily effected by mechanical factor in case of high stress(302MPa), but by electro-chemical reaction in a lower stress(137MPa). As stress level got lower and hardness of the 2nd phase got higher, the roughness of fracture surface increased.

• Nuclear Engineering and Technology
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• 제47권4호
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• pp.454-460
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• 2015

Acoustic emission (AE) is one of the promising methods for detecting the formation of stress corrosion cracks (SCCs) in laboratory tests. This method has the advantage of online inspection. Some studies have been conducted to investigate the characteristics of AE parameters during SCC propagation. However, it is difficult to classify the distinct features of SCC behavior. Because the previous studies were performed on slow strain rate test or compact tension specimens, it is difficult to make certain correlations between AE signals and actual SCC behavior in real tube-type specimens. In this study, the specimen was a AISI 304 stainless steel tube widely applied in the nuclear industry, and an accelerated test was conducted at high temperature and pressure with a corrosive environmental condition. The study result indicated that intense AE signals were mainly detected in the elastic deformation region, and a good correlation was observed between AE activity and crack growth. By contrast, the behavior of accumulated counts was divided into four regions. According to the waveform analysis, a specific waveform pattern was observed during SCC development. It is suggested that AE can be used to detect and monitor SCC initiation and propagation in actual tubes.

• 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.

• International Journal of Concrete Structures and Materials
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• 제10권sup3호
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• pp.43-52
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• 2016

The degradation of the load-bearing capacity of reinforced concrete beams due to corrosion has a profoundly negative impact on the structural safety and integrity of a structure. The literature is limited with regard to models of bond characteristics that relate to the reinforcement corrosion percentage. In this study, uniaxial tensile tests were conducted on specimens with irregular corrosion of their reinforced concrete. The development of cracks in the corroded area was found to be dependent on the level of corrosion, and transverse cracks developed due to tensile loading. Based on this crack development, the average stress versus deformation in the rebar and concrete could be determined experimentally and numerically. The results, determined via finite element analysis, were calibrated using the experimental results. In addition, bond elements for reinforced concrete with corrosion are proposed in this paper along with a relationship between the shear stiffness and corrosion level of rebar.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.108-113
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• 2007

In this report, stress corrosion cracking generation due to pipe material degradation in the primary stage of the nuclear power plant was investigated. Firstly, after artificially degrading the CF8A steel during 2, 4, and 6 months in actual temperature, $400^{\circ}C,$ assessed corrosion susceptibility of the degraded material following ASTM G5 standard. And next, the S.C.C. tests for the degraded material were conducted under the condition of $60^{\circ}C,$ 2wt.% H2BO3+Li70H solution, 0.8 oy. From the results, Corrosion rates linearly increased with degradation period and solution temperature increase. And both the raw material and the degraded materials were not failed in the S.C.C. test condition. In spite of long time test (about 3,900 hrs) under S.C.C. condition, surface pits or surface corrosion by the electro chemical reaction were not observed. And also, even though the nondestructive DCPD and ACPD methods were applied to on-line monitor the S.C.C. failure processes it was impossible because the surface pits and cracks were not generated.

• Corrosion Science and Technology
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• 제20권4호
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• pp.210-229
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• 2021

To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.

• Journal of Welding and Joining
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• 제29권2호
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• pp.34-39
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• 2011

Nickel-based austenitic alloys such as Alloy 82 and 182 had been employed as the weld metals in nuclear power plants (NPPs) due to their high corrosion resistance as well as good mechanical properties. However, since the 2000s, the occurrence of primary water stress corrosion cracking has been reported in conjunction with these alloys in domestic and oversea NPPs. In the present work, we assumed an imaginary crack at the inner surface of a surge nozzle weld that had previously experienced the outside repair welding, and constructed its finite element model. Finite element analysis was performed with respect to the heat transfer, and then to the residual stress for obtaining the total applied stress distributions. These stress distributions were finally converted to the stress intensity factors for estimating crack growth rate. From the comparison of crack growth rate curves for the cases of no repair welding and outside repair welding, it was found that the outside repair welding did not exhibit negative effect on the crack growth for the surge nozzle under consideration in this work; in both cases, the cracks stopped growing before they became the through-wall cracks.

• 한국압력기기공학회 논문집
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• 제8권3호
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.