• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1738-1748
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• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• Journal of Ocean Engineering and Technology
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• v.7 no.2
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• pp.91-102
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• 1993

The elastic-plastic fracture thoughness J sub(IC) of Nuclear Reactor Vessel Steel(SA 508-3) which has high toughness was discussed at temperatures RT, $-20^{\circ}C$, $200^{\circ}C$ and 1/2/CT specimen was used for this study. Especially the two methods recommended in ASTM and JSME were compared. It was difficult to find J sub(IC) by ASTM R-curve method with the specimen used for this research, while JSME R-curve method yielded good result. The stretched zone width menthod gave slightly larger J sub(IC) values than those by the R-curve method for SA 508-3 steel.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2003.10a
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• pp.146-151
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• 2003

Reduced Activation Ferritic/Martenstic (RAFs) are leading candidates for structural materials of D-T fusion reactor. One of The RAFs, JLF-1 (9Cr-2W-V, Ta) has been developed and proved to have good resistance against high-fluency neutrino irradiation and good phase stability. Recently, in order to clarify the strengthening mechanical at high temperature, a new scheme to improve high temperature mechanical properties is desired. Therefore, the creep properties and creep life prediction by Larson-Miller Parameter method for JLF-1 to be used for fusion reactor materials or other high temperature components were presented at the elevated temperatures of $500^{\circ}C$, $550^{\circ}C$, $600^{\circ}C$, $650^{\circ}C$ and $704^{\circ}C$. It was confirmed experimentally and quantitatively that a creep life predictive e벼ation at such various high temperatures was well derived by LMP.

• Journal of Ocean Engineering and Technology
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• v.18 no.2
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• pp.58-63
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• 2004

Reduced Activation Ferritic/Martensitic Steels (RAFs) are leading candidntes for structural materials of a D-T fusion reactor. One of the RAFs, JLF-l (9Cr-2W-V, Ta) has been developed and has shown to have good resistance against high-fluency neutrino irradiation and good phase stability. Recently, in order to clarify the strengthening mechanisms at high temperatures, a new scheme to improve high temperature mechanical properties is desired. Therefore, the test technique development of high temperature creep behaviors for this material is very important. In this paper, the creep properties and creep life prediction, using the Larson-Miler parameter method for JLF-l to be used for fusion reactor materials or other high temperature components, are presented at the elevated temperatures of 50$0^{\circ}C$, 55$0^{\circ}C$, $600^{\circ}C$, $650^{\circ}C$ and 704$^{\circ}C$. It was confirmed, experimentally and quantitatively, that a creep life predictive equation, at such various high temperatures, is well derived mr the LMP method.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3635-3642
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• 2021

Natural circulation systems (NCSs) are extensively applied in nuclear power plants because of their simplicity and inherent safety features. For some passive natural circulation systems in floating nuclear power plants (FNPPs), the ocean is commonly used as the heat sink. Condensation induced water hammer (CIWH) events may appear as the steam directly contacts the subcooled seawater, which seriously threatens the safe operation and integrity of the NCSs. Nevertheless, the research on the formation mechanisms of CIWH is insufficient, especially in NCSs. In this paper, the characteristics of flow rate and fluid temperature are emphatically analyzed. Then the formation types of CIWH are identified by visualization method. The experimental results reveal that due to the different size and formation periods of steam slugs, the flow rate presents continuous and irregular oscillation. The fluid in the horizontal hot pipe section near the water tank is always subcooled due to the reverse flow phenomenon. Moreover, the transition from stratified flow to slug flow can cause CIWH and enhance flow instability. Three types of formation mechanisms of CIWH, including the Kelvin-Helmholtz instability, the interaction of solitary wave and interface wave, and the pressure wave induced by CIWH, are obtained by identifying 67 CIWH events.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Journal of Ocean Engineering and Technology
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• v.15 no.2
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• pp.88-93
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• 2001

In this paper, joints of Cu-1Cr-0.1Zr alloy to STS316L were performed by friction welding method. Particularly, Cu-1Cr-0.1Zr alloy is attractive candidate as nuclear power plant material and exibit the best combination of high strength and good electrical and thermal conductivity of any copper alloy examined. The stainless steel is a structural material while copper alloy acts as a heat sink material for the surface heat flux in the first wall. So, in this paper, not only the development of optimizing of friction welding with more reliability and more applicability but also the development of in-process real-time weld quality (such as strength and toughness) evaluation technique by acoustic emission for friction welding of such nuclear reactor component of Cu-1Cr-0.1Zr alloy to STS316L steel sere performed.

• 한국태양에너지학회:학술대회논문집
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• 2011.04a
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• pp.204-209
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• 2011

According to the environment report of UN, korea was classified as potable water shortage countries. Approximately 71% of the Earth's surface is covered by ocean. However, it is difficult to use for industry of residential purpose without a certain processing. The development of solar and waste-heat used absorption desalination technology have been examined as a viable option for supplying clean energy. In this study, the modelling of the main devices for solar and waste-heat used and adsorption desalination system was introduced. The design is divided into three parts. First, the evaporator for the vaporization of the top water is designed, and then the reactor for the adsorption and release of the steam is designed, followed by the condenser for the condensation of the fresh water is designed. In addition, new features based on the energy balance are also included to design absorption desalination system. In this basicresearch, One-bed(reactor) adsorption desalination plant that employ a low-temperature solar and waste energy was proposed and experimentally studied. The specific water yield is measured experimentally with respect to the time controlling parameters such as heat source temperatures, coolant temperatures, system switching and half-cycle operational times.

• Journal of Ocean Engineering and Technology
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• v.12 no.1
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• pp.50-57
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• 1998

The leak before break(LBB) concept is generalized on the design of LNG tanks, pressure vessels and nuclear reactor in that any leakage of containment, in whatever amount, will not result in catastropic failure. For this purpose it is necessary to determine the surface crack shape, the opening displacement and the risk of catastropic brittle fracture when it becomes a through crack. In this study the crack propagation behavior of surface flaws and the crack opening displacement of through cracks under combined membrane and bending stresses were investigated with fatigue tests and fracture toughness test of aluminium alloy A5083-O. And fracture mechanics analysis of the crack opening displacement of through cracks were made in order to develop a new model expressing the behaviors of COD under combined membrane and bending stresses.

• Journal of Welding and Joining
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• v.33 no.4
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• pp.1-6
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• 2015

Advances in laser technology have yielded a multitude of innovative processes and applications in various industries. Laser peening is a typical example invented in the mid-1990s as a surface technology, which converted residual stress from tension to compression by just irradiating successive laser pulses to metallic materials under aqueous environment without any surface preparation. The effects of laser peening have been experimentally studied on residual stress, stress corrosion cracking(SCC) susceptibility and fatigue properties with water-penetrable frequency-doubled Nd:YAG laser. In addition, laser peening has been widely used in aviation and aerospace industries, automobile manufacturing and nuclear plant. One of the most important causes to improve the above-mentioned properties is the deeper compressive residual stress induced by laser peening. Taking advantage of the process without reacting force against laser irradiation, a remote operating system was developed to apply laser peening to nuclear power reactors as a preventive maintenance measure against SCC.