• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.4
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• pp.374-382
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• 2009

Steam generator(SG) tube, as a barrier isolating primary to the secondary coolant system of nuclear power plants(NPP), must maintain the structural integrity far the public safety and its efficient power generation capacity. And SG tubes bearing defects must be timely detected and taken repair measures if needed. For the accomplishment of these objectives, SG tubes have been periodically examined by eddy current testing(ECT) on the basis of administrative notices and intensified SG management program(SGMP). Stress corrosion cracking(SCC) on the SG tubes is not easily detected and even missed since it has lower signal amplitude and other disturbing factors against its detection. However once SCC is developed, that can cause detrimental affects to the SG tubes due to its rapid propagation rate. Accordingly SCC is categorized as prime damage mechanism challenging the soundness of the SG tubes. In this study, reproduced EDM notch specimens are examined for the detectability and quantitative characterization of the axial ODSCC by +PT MRPC probe, containing pancake, +PT and shielded pancake coils apart in a single plane around the circumference. The results of this study are assumed to be applicable fur providing key information of engineering evaluation of SCC and improvement of confidence level of ECT on SG tubes.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.137-144
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• 2009

In this paper, the signal analysis of eddy current array probe was performed to analyze the electromagnetic characteristics with the variation of FBH(flat bottomed hole) defects size on steam generator tube in NPP(nuclear power plants) using the electromagnetic finite element method. To obtain the electromagnetic characteristic of probes, the governing equation was derived from Maxwell's equations, and the individual problem was analyzed by using the 3-dimensional finite element method. For the simulation FBH defects were used. The depth of FBH defects were 20%, 40%, 60%, 80% and 100% of steam generator(SG) tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequencies of 100 kHz, 300 kHz and 400 kHz were used. Material of specimen was Inconel 600 which is usually used for SG tubes in NPP. The signal difference could be observed according to the size variation of depth of FBH defects and operation frequencies. The results in this paper can be helpful when the ECT(eddy current testing) signals from EC array probe are evaluated and analyzed.

• Journal of the Korean Society for Nondestructive Testing
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• v.27 no.5
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• pp.375-382
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• 2007

Nondestructive testing (NDT) by using the electromagnetic methods are useful for detecting cracks on the surface and subsurface of the metal. However, when the material contains both ferromagnetic and paramagnetic materials, it is difficult for NDT to detect and analyze cracks using this method. In addition the existence of a partial ferromagnetic material can be incorrectly characterized as a crack in the several cases. On the other hand a large crack has sometimes been misunderstood as a partially magnetized region. Inconel 600 is an important material in atomic energy plant. A nickel film is coated when a crack a appears on an Inconel substrate. Cracks are difficult to detect on the combined material of an Inconel substrate with a nickel film, which are paramagnetic and ferromagnetic material respectively. In this paper, a scan type magnetic camera, which uses a complex induced current-magnetic flux leakage (CIC-MFL) method as a magnetic source and a linearly integrated Hall sensor array (LIHaS) on a wafer as the magnetic sensors, was examined for its ability to detect cracks on the combined material. The evaluation probability of a crack is discussed. In addition the detection probability of the minimum depth was reported.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.5
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• pp.436-442
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• 2008

Graphite material has been recognized as a very competitive candidate for reflector, moderator, and structural material for very high temperature reactor (VHTR). Since VHTR is operated up to $900-950^{\circ}C$, small amount of impurity may accelerate the oxidation and degradation of carbon graphite, which results in increased porosity and lowered fracture toughness. In this study, ultrasonic wave propagation properties were investigated for both as-received and degradated material, and the feasibility of ultrasonic testing (UT) was estimated based on the result of ultrasonic property measurements. The ultrasonic properties of carbon graphite were half, more than 5 times, and 1/3 for velocity, attenuation, and signal-to-noise (S/N) ratio respectively. Degradation reduces the ultrasonic velocity slightly by 100 m/s, however the attenuation is about 2 times of as-receive state. The results of probability of detection (POD) estimation based on S/N ratio for side-drilled-hole (SDHs) of which depths were less than 100 mm were merely affected by oxidation and degradation. This result suggests that UT would be reliable method for nondestructive testing of carbon graphite material of which thickness is not over 100 mm. In accordance with the result produced by commercial automated ultrasonic testing (AUT) system, human error of ultrasonic testing is barely expected for the material of which thickness is not over 80 mm.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.149-157
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• 2005

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered at inside surface of hot cell and surface of equipment in hot cell. It was necessary to develop a particulate filtration equipment to reuse PFC solution used on PFC decontamination due to its high cost and to minimize the volume of second wastewater. Contamination characteristics of hot particulate were investigated and then a filtration process was presented to remove hot particulate in PFC solution generated through PFC decontamination process. The removal efficiency of PVDF(Poly vinylidene fluoride), PP(Polypropylene), Ceramic(Al$_{2}$O$_{3}$ filter showed more than 95$\%$. The removal efficiency of PVDF filter was a little lower than those of other kiters at same pressure(3psi). A ceramic filter showed a higher removal efficiency with other filters, while a little lower flux rate than other filters. Due to inorganic composition, a ceramic filter was highly stable against radio nuclides in comparison with PVDF and PP membrane, which generate H$_{2}$ gas in e-radioactivity atmosphere. Therefore, the adoption of ceramic filter is estimated to be suitable for the real nitration process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.77-85
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• 2006

A demonstration facility for an advanced spent fuel conditioning process (ACP) is under construction at KAERI. In this hotcell facility, the rear door is frequently used since all process equipment and materials are taken in and out only through the rear door. Therefore , both the structural safety and stability of the door are essentially required for the safety of ACP facility. In this paper, the finite element analysis has been performed to investigate the structural safety under the impact condition between the rear door and the door frame. Also the possibility of the rear door being tumbled over by the impact force or the inertia force under a sudden stop conditon has been evaluated. The analysis results demonstrate that the structural safety and stability of the rear door are sufficiently assured for both the impact and the accidential stop conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.17-24
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• 2006

A demonstration facility for advanced spent fuel conditioning process (ACP) is under construction in KAERI. In this hot cell facility, all process equipments and materials are taken in and out only through the rear door. The working table in front of the process rear door is specially designed to be openable for the efficient use of the space. This paper presents the structural safety analysis of the openable working table, for the normal operational load condition and accidential drop condition of heavy object. Both cases are investigated through static and dynamic finite element analyses. The analysis results show that structural safety of the working table is sufficiently assured and the working table is not collapsed even when an object of 500 kg is dropped from the height of 50 cm.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.17-23
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• 2008

In this study, plasma processing of metal surface is experimentally investigated to enhance the surface decontamination efficiency and to find out the reaction mechanism. Cobalt, the major contaminant in the nuclear facilities, and three fluorine-containing gases, $CF_4/O_2$, $SF_6/O_2$, and $NF_3$ are chosen for the investigation. Thin metallic disk specimens are prepared and their surface etching reactions with the three plasma gases are examined. Results show that the maximum etching rate of $17.2\;{\mu}m/min.$ is obtained with NF3 gas at $420^{\circ}C$, while with $CF_4/O_2$, $SF_6/O_2$ gas plasmas those of $2.56\;{\mu}m/min.$ and $1.14\;{\mu}m/min.$ are obtained, respectively. Along with etching experiments, constituent elements of the reaction products are identified to be cobalt, oxygen, and fluorine by AES (Auger Electron Spectroscopy) analysis. It turns out that the oxygen atoms are physically adsorbed ones to the surface from the ambient not participation ones during the analysis after reaction, which supports that the surface reaction of cobalt is mainly to be a fluorination reaction.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.9-17
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• 2007

In order to compare the correlation of radioactivity ratio between the radwaste streams and the primary coolant of PWR NPPs, A RCS sampling kit was installed to primary coolant system for the collection of the radionuclides during the normal operation of NPPs. RCS samples were collected from PWR type of domestic NPPs through 2004 to 2005, and pretreated with acid microwave digestion or leaching method to assay quantitatively of several interesting radionuclides. The radioactivity ratios of $^{137}Cs\;to\;^{60}Co$ in a filter cartridge and a resin cartridge were 2.3E-2 and 7.3E-1, respectively. At a same period of the reactor operating cycle, the radioactivity ratios of $^{137}Cs\;to\;^{60}Co$ were 6.3E-1 for a evaporator bottom, 6.7E-1 for a spent resin, and 5.6E-2 for a dry active waste, so that these radwaste streams were identified as having similar characteristics with the corresponding RCS samples.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.257-265
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• 2007

Spent filter wastes of about 1,000 units (200 L) have been stored in the waste storage facility of the Korea Atomic Energy Research Institute since its operation. At the moment, to secure space in a waste storage facility as well as to efficiently manage spent filter wastes, it is necessary to conduct a compaction treatment of these spent filters, and finally, to repack the compacted spent filters into a 200 liter drum. To do that, the spent filter wastes were first classified according to their generation facilities, their generation date and their surface dose rate by investigating the inventory of the spent filters. In order to repack a compacted spent filter in a 200 liter drum, it is first necessary to conduct a radionuclide assessment of a spent filter before compacting it. Therefore, after taking a representative sample from a spent filter without a dismantlement, the nuclide analysis for it will be conducted. And then, after putting a spent filter into a regular drum by conducting the columnar shaping of the hexahedral form of a spent filter, the compaction treatment of the shaped spent filter will be conducted by vertically compacting it.