• 한국수자원학회:학술대회논문집
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• 한국수자원학회 2017년도 학술발표회
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• pp.265-269
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• 2017

Because of abrupt changes for velocity in water, transient flow is occurred in practical life. To reduce and avoid the high or low pressure of pipe network system, various surge protection facilities are used to prevent the risk in pipe network system. Especially, we focused on study not only preventing positive and negative pressure but also selecting adequate equipment for high pressurized pipelines. Several critical cases were considered by undertaking a steady state hydraulic study and transient dynamic simulation and we suggested that the surge vessel of various surge protection system was recommended to control high and low pressures on pipeline system in perspective.

• Structural Engineering and Mechanics
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• 제52권6호
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• pp.1135-1156
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• 2014

The ratcheting effect greatly challenges the design of piping components. With the assistance of the quasi-three point bending apparatus, ratcheting and the ratcheting boundary of pressurized straight Z2CND18.12N stainless steel pipe under bending loading and vertical displacement control were studied experimentally. The characteristics of progressive inelastic deformation in axial and hoop directions of the Z2CND18.12N stainless steel pipes were investigated. The experiment results show that the ratcheting strain occurs mainly in the hoop direction while there is less ratcheting strain in the axial direction. The characteristics of the bending ratcheting behavior of the pressure pipes were derived and compared under load control and displacement control, respectively. The results show that the cyclic bending loading and the internal pressure affect the ratcheting behavior of the pressurized straight pipe significantly under load control. In the meantime, the ratcheting characteristics are also highly associated with the cyclic displacement and the internal pressure under displacement control. All these factors affect not only the saturation of the ratcheting strain but the ratcheting strain rate. A series of multi-step bending ratcheting experiments were conducted under both control modes. It was found that the hardening effect of Z2CND18.12N stainless steel pipe under previous cyclic loadings no matter with high or low displacement amplitudes is significant, and the prior loading histories greatly retard the ratcheting strain and its rate under subsequent loadings. Finally, the ratcheting boundaries of the pressurized straight Z2CND18.12N stainless steel pipe were determined and compared based on KTA/ASME, RCC-MR and the experimental results.

• Corrosion Science and Technology
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• 제4권3호
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• pp.108-113
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• 2005

One of the recent safety issues in the pressurized heavy water reactor (PHWR) is the cracking of the feeder pipe. Because of the limited accessibility to the cracked region and a high dose of radiation exposure, it is difficult to inspect all the pipes with the conventional ultrasonic method. In order to solve this problem, a long-range guided wave technique has been developed. A computer program to calculate the dispersion curves in the pipe was developed and the dispersion curves for the feeder pipes in PHWR plants were determined. Several longitudinal and/or flexural modes were selected from the review of the dispersion curves and an actual experiment has been carried out with the specific alignment of the piezoelectric ultrasonic transducers. They were confirmed as L(0,1)) and/or flexural modes(F(m,2)) by the short time Fourier transformation(STFT) and were sensitive to the circumferential cracks, but not to the axial cracks in the pipe. An electromagnetic acoustic transducers(EMAT) was designed and fabricated for the generation and reception of the torsional guided wave. The axial cracks were detected by a torsional mode(T(0,1)) generated by the EMAT.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2552-2564
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• 2020

The crevice stress corrosion cracking (SCC) susceptibility of austenitic stainless steels was evaluated in simulated pressurized water reactor (PWR) environments. To simulate the abnormal condition in temporary clamping devices on leaking small bore pipes, crevice bent beam (CBB) tests were performed in the oxygenated as well as hydrogenated conditions. No SCC cracks were found for SS316 in both conditions. SS304 also showed good resistance in the hydrogenated condition. However, all SS304 specimens showed SCC cracks in the oxygenated condition, indicating poor crevice SCC resistance. It was found that residual ferrites were selectively dissolved because of the galvanic corrosion coupled with the neigh-bouring austenite phase, resulting in SCC initiation in SS304. Crack morphologies were mostly transgranular assisted by the damaged δ-ferrite and deformation-induced slip bands.

• Corrosion Science and Technology
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• 제18권4호
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• pp.138-147
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• 2019

Pipes and components of the secondary system in the pressurized water reactor (PWR) are mainly comprised of manufactured carbon steel. Thus, the generated carbon steel corrosion products are transported into the steam generator and deposited, thereby deteriorating the integrity of the steam generator. Environmental condition in the secondary system of the PWRs differs across different locations. So, the corrosion rate and types of corrosion products depend on specific locations in the secondary system. In this study, the quantity and chemical compositions of corrosion products generated in various locations that vary in different temperatures and chemistry conditions were investigated. As a result of evaluating the PWR "Unit A" that is in current operation, the amount of corrosion products generated in the section of high temperature feedwater system was identified as the largest source in the secondary system. Major components of corrosion products were iron oxides such as magnetite, hematite, and lepidocrocite.

• 대한기계학회논문집A
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• 제33권2호
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.