• Journal of the Korean Society for Nondestructive Testing
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• v.37 no.1
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• pp.13-20
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• 2017

A conventional phased array ultrasonic system offers the ability to steer an ultrasonic beam by applying independent time delays of individual elements in the array and produce an ultrasonic image. In contrast, full matrix capture (FMC) is a data acquisition process that collects a complete matrix of A-scans from every possible independent transmit-receive combination in a phased array transducer and makes it possible to reconstruct various images that cannot be produced by conventional phased array with the post processing as well as images equivalent to a conventional phased array image. In this paper, a basic algorithm based on the LLL mode total focusing method (TFM) that can image crack type flaws is described. And this technique was applied to reconstruct flaw images from the FMC dataset obtained from the experiments and ultrasonic simulation.

• Journal of the Korean Society for Nondestructive Testing
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• v.37 no.1
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• pp.21-28
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• 2017

Eddy current array (ECA) technology provides the ability to electronically scan without mechanical scanning for a footprint of probe that consists of several eddy current coils arranged side-by-side and two dimensionally. Compared to single-coil eddy current technology, the ECA technology has a higher inspection speed and reliability because a large area can be covered in a single-probe pass and the obtained images can facilitate data interpretation. In this study, we developed an eddy current array probe with $2{\times}16$ coil arrays and a multiplexer that can select transmit-receive coils sequentially and operate in transmit-receive mode. Surface inspection was performed using this probe for specimens that had various flaws and the characteristics of transmit-receive mode ECA technology was studied through an analysis of the obtained eddy current signals and C-scan images.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.457-463
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• 2005

KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr.

• Journal of Energy Engineering
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• v.28 no.4
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• pp.103-110
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• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.

• Corrosion Science and Technology
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• v.15 no.2
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• pp.84-91
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• 2016

The transport and deposition of corrosion products in pressurized water nuclear reactor (PWR) steam generators have led to corrosion (SCC, denting etc.) problems. Lancing, mechanical cleaning and chemical cleaning have been used to reduce these problems. The methods of lancing and mechanical cleaning have limitations in removing corrosion products due to the structure of steam generator tubes. But high temperature chemical cleaning (HTCC) with EDTA is the most effective method to remove corrosion products regardless of the structure. However, EDTA in chemical cleaning aqueous solution and chemical cleaning environments affects the integrity of materials used in steam generators. The nuclear power plants have to perform the pre-test (also called as qualification test (QT)) that confirms the effect on the integrity of materials after HTCC. This is one of the series studies that assess the effect, and this study determines the effects of 20 % EDTA aqueous solution on defective tubes in high temperature chemical cleaning environments. The depth and magnitude of defects in steam generator (SG) tubes were measured by eddy current test (ECT) signals. Surface analysis and magnitude of defects were performed by using SEM/EDS. Corrosion rate was assessed by weight loss of specimens. The ECT signals (potential and depth %) of defective tubes increased marginally. But the lengths of defects, oxides on the surface and weights of specimens did not change. The average corrosion rate of standard corrosion specimens was negligible. But the surfaces on specimens showed traces of etching. The depth of etching showed a range on the nanometer. After comprehensive evaluation of all the results, it is concluded that 20 % EDTA aqueous solution in high temperature chemical cleaning environments does not have a negative effect on defective tubes.

• The Journal of Korean Institute of Electromagnetic Engineering and Science
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• v.23 no.11
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• pp.1315-1322
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• 2012

Frequency domain reflectometry diagnoses faults on electric cables by measuring the cable impedance. Time domain impedance measurement technique using an oscilloscope instead of a network analyzer is widely used for electric power cables under harsh environment or powered condition. However, impedance measurement in the time domain shows inaccuracy as the frequency increases due to several parasitic impedances, which results in the poor resolution of fault points. This paper presents the accuracy enhancement technique using a module with an operational amplifier and an error calibration method in the time domain impedance measurements, which is confirmed by comparing the cable impedance measurement results.

• Transactions of the Korean hydrogen and new energy society
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• v.23 no.3
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• pp.219-226
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• 2012

A nuclear fusion fuel cycle plant is composed of various subsystems such as a hydrogen isotope storage and delivery system, a tokamak exhaust processing system, and a hydrogen isotope separation system. Korea shares in the construction of the International Thermonuclear Experimental Reactor fuel cycle plant with the EU, Japan and US, and is responsible for the development and supply of the storage and delivery system. We thus present details on the hydrogen isotope process safety. The main safety analysis procedure is to use a hazard and operability study. Nine segments were studied how the plant might deviate from its design purpose. We present a detailed description of the process, examine every part of it to determine how deviations from the design intent can occur and decide whether these deviations can give rise to hazards. We determine possible causes and note protective systems, evaluate the consequences of the deviation, and recommend actions to achieve our safety goal.

• Corrosion Science and Technology
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• v.13 no.3
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• pp.112-119
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• 2014

In this work, it is aimed to develop the fabrication method of axial stress corrosion cracking (SCC) defects having various sizes, on the outer diameter surface of the steam generator (SG) tubings. To control the length of the artificial SCC defect, the specific area of the SG tubing samples was exposed to an acidic solution after a sensitization heat treatment. During the exposure to an acidic solution, a direct current potential drop (DCPD) method was adopted to monitor the crack depth. The size of the SCC defect was first evaluated by an eddy current test (ECT), and then confirmed by a destructive examination. From the comparison, it was found that the actual crack length was well controlled to be similar to the length of the surface exposed to an acidic solution (5, 10, 20 or 30 mm in this work) with small standard deviation. From in-situ monitoring of the crack depth using the DCPD method, it was possible to distinguish a non-through wall crack from a through wall crack, even though the depth of the non-through wall crack was not able to be precisely controlled. The fabrication method established in this work was useful to simulate the SCC defect having similar size and ECT signals as compared to the field cracks in the SG tubings of the operating Korean PWRs.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.710-720
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• 2017

In Korea, many nuclear power plants operate at a single site based on geographical characteristics, but the population density near the sites is higher than that in other countries. Thus, multiunit accidents are a more important consideration than in other countries and should be addressed appropriately. Currently, there are many issues related to a multiunit probabilistic safety assessment (PSA). One of them is the quantification of a multiunit PSA model. A traditional PSA uses a Boolean manipulation of the fault tree in terms of the minimal cut set. However, such methods have some limitations when rare event approximations cannot be used effectively or a very small truncation limit should be applied to identify accident sequence combinations for a multiunit site. In particular, it is well known that seismic risk in terms of core damage frequency can be overestimated because there are many events that have a high failure probability. In this study, we propose a quantification method based on a Monte Carlo approach for a multiunit PSA model. This method can consider all possible accident sequence combinations in a multiunit site and calculate a more exact value for events that have a high failure probability. An example model for six identical units at a site was also developed and quantified to confirm the applicability of the proposed method.

• Journal of the Korea institute for structural maintenance and inspection
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• v.12 no.2
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• pp.75-82
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• 2008

There are almost 18,000 dams including about 1,200 large dams in Korea. The large dams are well operated and maintained by KWATER(Korea Water Resource Corporation), KRC(Korea Rural Community & Agriculture Corporation) and KHNP(Korea Hydro & Nuclear Power Co., Ltd.). Several research reports concern with the safety of these large dams are presented but there is no paper concerned with small dams which has diversion tunnel through the dam body. The purpose of this study is to show the common defects of small dams according to various cases of degradation of dams and the repair and retrofit methods which applied to the damaged dams. And this study performed resistivity survey to evaluate the effect of retrofitting dam. Also, this study tries to present the solution which concerned with these common defects in maintenance and design steps.