• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.186-192
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• 2022

Flow accelerated corrosion (FAC) is a type of pipe corrosion in which the pipe thickness decreases depending on the fluid flow conditions. In nuclear power plants, FAC mainly occurs in the carbon steel pipes of a secondary system. However, because the temperature of a secondary system pipe is over 150 ℃, in situ monitoring using a conventional ultrasonic non-destructive testing method is difficult. In our previous study, we developed a waveguide ultrasonic thickness measurement system. In this study, we applied a waveguide ultrasonic thickness measurement system to monitor the thinning of the pipe according to the change in pH. The Korea Atomic Energy Research Institute installed FAC-proof facilities, enabling the monitoring of internal fluid flow conditions, which were fixed for ~1000 h to analyze the effect of the pH. The measurement system operated without failure for ~3000 h and the pipe thickness was found to be reduced by ~10% at pH 9 compared to that at pH 7. The thickness of the pipe was measured using a microscope after the experiment, and the reliability of the system was confirmed with less than 1% error. This technology is expected to also be applicable to the thickness-reduction monitoring of other high-temperature materials.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.417-422
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• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.5
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• pp.1106-1116
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• 1994

Leak Before Break(LBB) design concept is applied to piping systems of newly-built Yonggwang 3, 4 nuclear generating stations as a design alternative to the provision of pipe whip restraints, in recognition of the questionable benefits of providing such restraints. The objective of this paper is to evaluate the material properties (tensile and fracture toughness) of SA312 TP316 stainless steel and their associated welds manufactured for shutdown cooling system of Yonggwang 3, 4 nuclear generating stations. Effect of various parameters such as specimen orientation, test temperature, welding on material properties were examined.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2000.10a
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• pp.42-46
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• 2000

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

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1346-1353
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• 2017

The FISPACT-II inventory simulation platform is a modern computational tool with advanced and unique capabilities. It is sufficiently flexible and efficient to make it an ideal basis around which to perform extensive simulation studies to scope a variety of responses of many materials (elements) to several different neutron irradiation scenarios. This paper briefly presents the typical outputs from these scoping studies, which have been used to compile a suite of nuclear physics materials handbooks, providing a useful and vital resource for material selection and design studies. Several different global responses are extracted from these reports, allowing for comparisons between materials and between different irradiation conditions. A new graphical output format has been developed for the FISPACT-II platform to display these "global summaries"; results for different elements are shown in a periodic table layout, allowing side-by-side comparisons. Several examples of such plots are presented and discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.8-17
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• 2019

Perforated plates are used for the steam generator tube-sheet and the Reactor Vessel Closure Head in the Nuclear Power Plant. The ASME code, Section III Appendix A-8000, addresses the analysis of perforated plates, however, this analysis is only limited to the flat plate with a triangular perforation pattern. Based on the concept of the effective elastic constants, simulation of flat and spherical perforated plates and their equivalent solid plates were carried out using Finite Element Analysis (FEA). The isotropic material properties of the perforated plate were replaced with anisotropic material properties of the equivalent solid plate and subjected to the same loading conditions. The generated curves of effective elastic constants vs ligament efficiency for the flat perforated plate were in agreement with the design curve provided by ASME code. With this result, a plate with spherical curvature having perforations can be conveniently analyzed with equivalent elastic modulus and equivalent Poisson's ratio.

• The Korean Journal of Nuclear Medicine Technology
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• v.21 no.1
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• pp.60-64
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• 2017

Purpose Westgard multi-rules application based on test quality improvement and commercialized international standard has been widely used in quality control. However, it is difficult to applicate the Westgard multi-rules in nuclear medicine in vitro tests due to the larger sample sizes and the simultaneous measurement of quality control material and patient sample. This study investigated the usefulness of Westgard multi-rules application in nuclear medicine in vitro tests. Materials and Methods A total of 282 systematic error multi-rules (22s, 101s) recorded in the samsung medical center computer system from January 2013 to June 2016 along with 117 cases of corrective measure record was analyzed. The Quality control implementation is recorded in Hospital information system were divided into 4 high-level areas including quality control material error, experimental procedural error, Kit lot number management error, and others. To prevent quality control material error, the existing method that each staff used their own method was changed. The staff who in charge of managing the quality control material was designated and daily consumption amount of every test was strictly controlled by one person. To prevent other errors, every test step was standardized so that the entire test procedures are identically implemented. Results The total quality control implementation was 117 cases; As a result, 62 quality control material errors were 62 cases, experimental process errors were 24 cases, Kit lot number control errors were 18 cases, and other errors were 13 cases. The quality control material error was corrected and could be used fresh materials within 2 days after thawing. The cases of systemic error were decreased to causes as quality control material error. The quality control materials were reduced above 10 vials to a monthly average. In addition, these errors of experimental processing and Kit lot number were improved by test standardization. Consequently, the cases of 101s and 22s in systematic error rules decreased at least 2 cases to a monthly average. Conclusion To confirm of systematic error through multi-rules application quickly, it is necessary to base on management of the QC material, target values and standard deviation. Moreover, in the event of a systematic error, it was found important to record measures based on test cause analysis. The experiment results are expected to contribute to internal quality control improvement and prompt and accurate result reporting through error recording and causal analysis based on Westgard multi-rules analysis.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.820-830
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• 2016

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.332-339
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• 2015

The leak before break (LBB) concept is widely used in designing pipe lines in nuclear power plants. According to the concept, the amount of leaking liquid from a pipe should be more than the minimum detectable leak rate of a leak detection system before catastrophic failure occurs. Therefore, accurate estimation of the leak rate is important to evaluate the validity of the LBB concept in pipe line design. In this paper, a program was developed to estimate the leak rate through circumferential cracks in pipes in nuclear power plants using the Henry-Fauske flow model and modified Henry-Fauske flow model. By using the developed program, the leak rate was calculated for a circumferential crack in a sample pipe, and the effect of the flow model on the leak rate was examined. Treating the crack morphology parameters as random variables, the statistical behavior of the leak rate was also examined. As a result, it was found that the crack morphology parameters have a strong effect on the leak rate and the statistical behavior of the leak rate can be simulated using normally distributed crack morphology parameters.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1264-1272
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• 2016

The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB) concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry-Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.