• Research in Plant Disease
• /
• v.17 no.3
• /
• pp.364-368
• /
• 2011

Gray mold caused by Botrytis cinerea was found on a carrot seedling in a greenhouse and a field at Daegwallryeong, Gangwon Province in 2007-2009. Symptoms included irregular, brown, blight, or chlorotic halo on leaves and petioles of the carrots. Fungal conidia were globose to subglobose or ellipsoid, hyaline or pale brown, nonseptate, one celled, $7.2-18.2{\times}4.5-11\;{\mu}m$ ($12.1{\times}8.3\;{\mu}m$) in size, and were formed on botryose heads. B. cinerea colonies were hyaline on PDA, and then turned gray and later changed dark gray or brown when spores appeared. The fungal growth stopped at $35^{\circ}C$, temperature range for proper growth was $15-25^{\circ}C$ on MEA and PDA. Carrots inoculated with $1{\times}10^5$ ml conidial suspension were incubated in a moist chamber at $25{\pm}1^{\circ}C$ for pathogenicity testing. Symptoms included irregular, brown, water-soaked rot on carrot roots and irregular, pale brown or dark brown, water-soaked rot on leaves. Symptoms were similar to the original symptoms under natural conditions. The pathogen was reisolated from diseased leaves, sliced roots, and whole roots after inoculation. As a result, this is the first report of carrot gray mold caused by B. cinerea in Korea.

• Journal of the Korean Institute of Gas
• /
• v.21 no.5
• /
• pp.89-94
• /
• 2017

The VR(Virtuality Reality) technology provides very close experience to reality by stimulating humans' external recognition with artificial technologies. In order to overcome the limitation of real-environment training, VR is being applied in industry field as a key technology to prevent safety accident and its control procedure training. However, it is difficult to build VR-based training system because 3D modeling and software coding are necessary for materialization of VR environment demands of many development resource. In this research referring to VR based training content implementation, a method to utilizing VRDC(VR-based Dynamic visualization Component) is suggested and by applying it to plant safety training system, it was confirmed its practicality.

• Food Science of Animal Resources
• /
• v.34 no.6
• /
• pp.763-768
• /
• 2014

The identity of 45 Hanwo and 47 imported beef (non-Hanwoo) samples from USA and Australia were verified using the microsatellite (MS) marker and single nucleotide polymorphism (SNP) methods. Samples were collected from 19 supermarkets located in the city of Seoul and Gyeonggi province, South Korea, from 2009 to 2011. As a result, we obtained a 100% concordance rate between the MS and SNP methods for identifying Hanwoo and non-Hanwoo beef. The MS method presented a 95% higher individual discriminating value for Hanwoo (97.8%) than for non-Hanwoo (61.7%) beef. For further comparison of the MS and SNP methods, blood samples were collected and tested from 54 Hanwoo ${\times}$ Holstein crossbred cattle (first, second, and third generations). By using the SNP and MS methods, we correctly identified all of the first-generation crossbred cattle as non-Hanwoo; in addition, among the second and third generation crossbreds, the ratio identified as Hanwoo was 20% and 10%, respectively. The MS method used in our study provides more information, but requires sophisticated techniques during each experimental process. By contrast, the SNP method is simple and has a lower error rate. Our results suggest that the MS and SNP methods are useful for discriminating Hanwoo from non-Hanwoo breeds.

• Corrosion Science and Technology
• /
• v.20 no.5
• /
• pp.308-314
• /
• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.26 no.5
• /
• pp.181-190
• /
• 2022

In assessing the seismic safety of nuclear power plants, it is essential to analyze the structures using the observed ground motion. In particular, spatial variation in which the characteristics of the ground motion record differ may occur if the location is different within the site and even if the same earthquake is experienced. This study analyzed the spatial variation characteristics of the ground motion observed at the structure and site using the earthquake records measured at the Hamaoka nuclear power plant. Even if they were located on the same floor within the same unit, there was a difference in response depending on the location. In addition, amplification was observed in Unit 5 compared to other units, which was due to the rock layer having a slower shear wave velocity than the surrounding bedrock. Significant differences were also found in the records of the structure's foundation and the free-field surface. Based on these results, the necessity of considering spatial variation in the observed records was suggested.

• Journal of the Korean Society of Safety
• /
• v.27 no.5
• /
• pp.229-234
• /
• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.56-64
• /
• 2020

The safety related heat exchangers have been evaluated for their performance during the operation of the nuclear power plant. The evaluation program for the safety related heat exchanger was developed in 2010 and used by KHNP based on EPRI TR-10739 algorithms. The spend fuel pool heat exchanger is one of the safety related heat exchanger in the nuclear power plant and also evaluated for their performance. Recently the performance evaluation for the spend fuel pool heat exchanger was not available because of the decreased heat in the spend fuel pool due to the long term overhaul. This paper analyzes the main cause of evaluation failure in the evaluation process and suggests the criteria for the heat exchanger performance evaluation during the long term overhaul.

• Journal of Applied Reliability
• /
• v.16 no.3
• /
• pp.231-237
• /
• 2016

Purpose: It is not easy to suggest a quantitative data related to safety analysis. The objective of this paper is to propose a method of Safety Integrity Level (SIL) analysis and to suggest a SIL analysis result for single station controller in nuclear power plant based on IEC 61508. Methods: The Failure Modes and Effects Diagnostic Analysis (FMEDA) and average probability of failure on demand (PFD) are used for SIL assessment. Results: A SIL of single station controller is evaluated 4 by a reliability analysis results and PFD. Conclusion: A SIL analysis method and result for single station controller based on IEC 61508 are proposed in this paper. It can applicable for a manufacturer data in safety-related system.

• Nuclear Engineering and Technology
• /
• v.42 no.4
• /
• pp.414-425
• /
• 2010

In this paper, a procedure is developed for identifying a number of representative solutions manageable for decision-making in a multiobjective optimization problem concerning the test intervals of the components of a safety system of a nuclear power plant. Pareto Front solutions are identified by a genetic algorithm and then clustered by subtractive clustering into "families". On the basis of the decision maker's preferences, each family is then synthetically represented by a "head of the family" solution. This is done by introducing a scoring system that ranks the solutions with respect to the different objectives: a fuzzy preference assignment is employed to this purpose. Level Diagrams are then used to represent, analyze and interpret the Pareto Fronts reduced to the head-of-the-family solutions.

• Nuclear Engineering and Technology
• /
• v.50 no.8
• /
• pp.1210-1216
• /
• 2018

Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.