• Proceedings of the KSME Conference
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• 2000.04a
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• pp.908-913
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• 2000

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 evaluations 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, two dimensional finite element analyses were applied for various surface crack. Total of 18 crack geometries were analyzed, and Q 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-tin stress field due to constraint effect.

• Journal of the Korean institute of surface engineering
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• v.52 no.6
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• pp.293-297
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• 2019

Quasi-nano-hardness and corrosion behaviors of neutron-irradiated reactor pressure vessel (RPV) steels such as 15Ch2MFA (Ni<0.4, 2.520 n/㎠ (En>1.0 MeV) for 32 days. Quasi-nano-hardnesses of the 15Ch2MFA and 15Cr2NHFA steels were 183.8 and 179.8 Hv, respectively. Their corrosion rates and corrosion potentials were 2.4×10^-4Acm^-2, -515.9 mV_SHE and 6.8×10^-4 Acm^-2, -523.6 mV_SHE in NACE standard TM0284-96 solution at room temperature, respectively. 15Ch2MFA steel showed better quasi-nano-hardness and corrosion resistance than 15Cr2NHFA steel in this test condition.

• Journal of Ocean Engineering and Technology
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• v.14 no.1
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• pp.44-51
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• 2000

Safety and integrity are required for reactor pressure vessels because they are operated in high temperature. There are single specimen method multiple specimen method and load ratio analysis method which used as evaluation of safety and integrity for reactor pressure vessels. In this study the fracture resistance curve(J-R curve) elastic-plastic fracture toughness($J_{IC}$) and material tearing modulus ($T_{mat}$) of SA 508 class 3 alloy steel used as reactor pressure vessel steel are measured and evaluated at room temperature 20$0^{\circ}C$ and 30$0^{\circ}C$ according to unloading compliance method and load ration analysis method. And then the comparison with experimental $J_{IC}$ and theoretical$J_{IC}$ by local fracture strain is managed.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.273-286
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• 2007

This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical $CO_2$, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.

• 전기의세계
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• v.19 no.4
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• pp.18-23
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• 1970

A neutron detector, in general, can not be utilized as the thermal neutron detecting chamber in the nuclear power reactor, especially P.W.R. due to the characteristics of high temperature, high pressure and high neutron flux in a reactor vessel. We have performed an experiment to detect the thermal neutrons at 400.deg. C and high flux of thermal neutron in a power reactor. Coating boron-10 on the aluminium plates by means of surface diffusion method at 600.deg. C for 5 hours in an electric furace, also we made a typical chamber which was compensated ionization chamber filled with free air as an ionization gas. It was checked the chamber characteristics in the TRIGA MARK-II Reactor at the power level from zero to 250KW. The chamber current showed a perfect linear increase to power increase. However, many variation of the measured current were observed within the power of 50KW.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.2
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.15-19
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• 2014

A Failure or degradation of reactor upper head penetration is a recurring problem due to long term operation at nuclear power plants. And a flaw in the reactor upper head penetration has caused unplanned plant shutdown for repair as well as high economic impact on the plants. Consequently, a detection of flaws is of the utmost importance. Prior to the replacement of reactor upper head penetration, some utilities have repaired the flaws of reactor upper head penetration generated by overlay weld. Until now, only the base metal in reactor upper head penetration has been inspected according to 10 CFR 50.55a and ASME code case N-729-1. Accordingly, it is difficult to detect manufacturing defects and repair defects in overlay weld. This paper presents a case study on the application of Time of Flight Diffraction technique for reactor head penetration mockup with artificial flaws in overlay weld. This study offers a way to understand the flaws detected in reactor upper head penetration overlay weld.

• Korean Chemical Engineering Research
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• v.46 no.1
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• pp.189-193
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• 2008

In this study, the effect of reaction conditions on the transesterification of soybean oil and methanol was investigated in a high-pressure-fixed-bed-reactor-system with zinc aluminate catalysts. Without catalysts, high-pressure-reaction at $300^{\circ}C$ and 1,200 psi brought 19% yields of methyl esters, which was caused by the approach of reaction condition to supercritical point of methanol. However, except the specific reaction condition, the yields in the reaction with no catalyst were very low below 4.5%. The zinc aluminate was prepared as catalyst by coprecipitation and characterized with $N_2$ gas adsorption/desorption and X-ray diffraction. With catalyst, the effect of the reaction parameters such as temperature, pressure, and molar ratio of reactants on biodiesel production was demonstrated. The higher temperature, pressure, and methanol molar ratio to soybean oil, the more yields of methyl esters. It was proved that among the reaction parameters, the reaction temperature be the most influential variable on methyl ester yields.

• Bulletin of the Korean Chemical Society
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• v.23 no.5
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• pp.674-678
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• 2002

The dehydrogenation of propane to propylene has been studied in an isothermal high-temperature shell-and-tube membrane reactor containing a Pd-coated ${\psi}$-Al2O3 membrane and a Pt/K/Sn/Al2O3 packed catalyst . A tubular Pd-coated ${\psi}$-Al2O3 membrane was prepared by an electroless plating method. This membrane showed high hydrogen to nitrogen permselectivities (PH2N2 = 10-50) at 400 $^{\circ}C$ and 500 $^{\circ}C$ with various transmembrane pressure drops. The employment of a membrane reactor in the dehydrogenation reaction, which selectively separates hydrogen from the reaction mixture along the reaction path, can greatly increase the conversion and enable operation of the reactor at lower temperatures. High hydrogen permselectivity has been confirmed as a key factor in determining the reactor performance of conversion enhancement.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.19 no.1
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• pp.121-133
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• 2020

Underwater cutting has a different mechanism than dry cutting, and there are more restrictions than benefits. Due to these constraints, research and development of underwater cutting has been very limited. At present, reactor dismantling is emerging as an important task worldwide, and reactor pressure containers, a key part of the reactor, are decommissioned based on underwater cutting. Reactor pressure containers are high-level radioactive waste, which is one of the main goals of today, such as to bridge the gap between environmental, safety, and cutting performance; hence, a process suitable for cutting should be applied. Therefore, many studies are being conducted on underwater cutting in connection with the dismantling of nuclear reactors in various areas in order to find appropriate processes. This paper first introduces the core technology of underwater cutting processes and discusses various processes. The emphasis is then placed on the adequacy of the reactor dismantling application. More specifically, we examine the suitability for the mechanical and thermal cutting processes, respectively, to find a solution suitable for dismantling a reactor. We discuss how each solution can sufficiently perform the specified functions at each stage of reactor dismantling and suggest that these processes can perform all of the work of underwater cutting.