• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.37 no.9
• /
• pp.855-862
• /
• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.39 no.11
• /
• pp.905-911
• /
• 2015

In this study, to examine the effect of a hole size change(smaller hole diameter) in the outer region of the lower-support-structure-bottom plate(LSSBP) on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.15. The predicted results were compared with those of the original LSSBP. Through these comparisons, it was concluded that a more uniform distribution of the mass flow rate at the core-inlet plane could be obtained by reducing the hole size in the outer region of the LSSBP. Therefore, from the nuclear regulatory perspective, design change of the hole pattern in the outer region of the LSSBP may be desirable in terms of improving both the mechanical integrity of the fuel assembly and the core thermal margin.

• Journal of the Korean Institute of Gas
• /
• v.17 no.2
• /
• pp.50-58
• /
• 2013

Various process piping usually carries out high flammable and explosible gas under high pressure and high temperature. Due to frequent change of design and structure it becomes more complicated and compactly located. The safety management level is relatively low since it is considered as simply designed component. In this study a safety assessment procedure is proposed for complicated piping system around a mixing drum in which natural gas and by-product gases were mixed. According to ASME code, pipe stress analysis was conducted for determining design margin at some key locations of the piping. These high stress locations can be used as major inspection points for managing the pipe integrity. Sensitivity analysis with outside temperature of the pipe and support constraint condition. Possible effect of hydroen gas to the pipe steel during the previous use of the by-product gas was also discussed.

• Nuclear Engineering and Technology
• /
• v.40 no.4
• /
• pp.255-268
• /
• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.22 no.1
• /
• pp.15-23
• /
• 2009

Masonry structures that are very strong in compression fail due to the instability of structural shape of geometry rather than the material stress limit. Considering such structural behavior, the use of the limit theorem that focuses on structural collapse mechanisms is more appropriate for the evaluation of the structural safety of stone voussoir arch bridges. This paper is to investigate structural performance of the stone arch bridges constructed using dry construction method in Korea based on the limit theorem and to exploit the result to develop a system for an structural safety margin. It is expected that this study will help us understand structural behavior of stone voussoir arch bridges in Korea. Also, it will provide a guideline to make engineering decision from the viewpoint of the maintenance of cultural heritages.

• Journal of Aerospace System Engineering
• /
• v.16 no.6
• /
• pp.106-113
• /
• 2022

The structural safety of the basic design model of the linear actuator for the individual blade pitch control of eVTOL personal aircraft was investigated. Stress analysis based on the finite element method was conducted, and the margin of safety was calculated to examine the structural safety under stall load conditions. Additionally, fatigue analysis was conducted to evaluate the fatigue life of the linear actuators under operating conditions. The load history with the blade pitch angle was calculated using multi-body dynamics analysis, and the static load analysis was used to obtain the stress distribution for the rated load. As a result, it was confirmed that the safety margins exceeded zero, and the fatigue lives of all linear actuator components exceeded 107 cycles, indicating a safe structural range.

• Journal of the Korean Geotechnical Society
• /
• v.40 no.3
• /
• pp.101-108
• /
• 2024

A system with the ability to recognize potential key blocks during tunnel construction by analyzing the rock face was developed in this study. This system predicts the formation of key blocks in advance and evaluates their safety factors. A laser scanner was used to collect a three-dimensional point cloud of the rock face, which was then utilized to model the excavation surface and derive the joint surfaces. Because joint surfaces have specific strikes and dip angles, the key blocks formed by these surfaces are deduced through iterative calculations, and the safety factor of each key block can be calculated accordingly. The model experiments confirmed the accuracy of the system's output in terms of the joint surface characteristics. By inputting the joint surface information, the calculated safety factors were compared with those from the existing commercial software, demonstrating stable calculation results within a 1% error margin.

• Nuclear Engineering and Technology
• /
• v.36 no.6
• /
• pp.528-539
• /
• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Toxicological Research
• /
• v.35 no.2
• /
• pp.103-117
• /
• 2019

The mixture of 5-chloro-2-methylisothiazol-3(2H)-one (CMIT) and 2-methylisothiazol-3(2H)-one (MIT), CMIT/MIT, is a preservative in cosmetics. CMIT/MIT is a highly effective preservative; however, it is also a commonly known skin sensitizer. Therefore, in the present study, a risk assessment for safety management of CMIT/MIT was conducted on products containing 0.0015% of CMIT/MIT, which is the maximum MIT level allowed in current products. The no observed adverse effect level (NOAEL) for CMIT/MIT was 2.8 mg/kg bw/day obtained from a two-generation reproductive toxicity test, and the skin sensitization toxicity standard value for CMIT/MIT, or the no expected sensitization induction level (NESIL), was $1.25{\mu}g/cm^2/day$ in humans. According to a calculation of body exposure to cosmetics use, the systemic exposure dosage (SED) was calculated as 0.00423 mg/kg bw/day when leave-on and rinse-off products were considered. Additionally, the consumer exposure level (CEL) amounted to $0.77512{\mu}g/cm^2/day$ for all representative cosmetics and $0.00584{\mu}g/cm^2/day$ for rinse-off products only. As a result, the non-cancer margin of safety (MOS) was calculated as 633, and CMIT/MIT was determined to be safe when all representative cosmetics were evaluated. In addition, the skin sensitization acceptable exposure level (AEL)/CEL was calculated as 0.00538 for all representative cosmetics and 2.14225 for rinse-off products; thus, CMIT/MIT was considered a skin sensitizer when all representative cosmetics were evaluated. Current regulations indicate that CMIT/MIT can only be used at concentrations 0.0015% or less and is prohibited from use in other cosmetics products. According to the results of this risk assessment, the CMIT/MIT regulatory values currently used in cosmetics are evaluated as appropriate.

• Asian Pacific Journal of Cancer Prevention
• /
• v.15 no.17
• /
• pp.7069-7077
• /
• 2014

Background: This retrospective study aimed to validate the safety and effectiveness of hepatectomy for huge hepatocellular carcinoma (HCC). Materials and Methods: Data of patients who underwent hepatectomy for HCC between January 2006 and December 2012 were reviewed. The patients were divided into three groups: huge HCC(${\geq}10cm$ in diameter), large HCC(${\geq}5$ but<10 cm in diameter) and small HCC(<5cm in diameter). Results: Characteristics of pre-operative patients in all three groups were homogeneously distributed except for alpha fetal protein (AFP)(p<0.001).The 30, 60, 90-day post-operative mortality rates were not different among the three groups (p=0.785, p=0.560, and p=0.549). Laboratory data at 1, 3, and 7 days after surgery also did not vary. The 5-year overall survival (OS) and 5-year disease-free survival (DFS) rates in the huge and large HCC groups were lower than that of the small HCC group (OS: 32.5% vs 36.3% vs 71.2%, p=0.000; DFS: 20.0% vs 24.8% vs 40.7%, p=0.039), but there was no difference between the huge and large HCC groups (OS: 32.5% vs 36.3%, p=0.667; DFS: 20.0% vs 24.8%, p=0.540). In multivariate analysis, five independent poor prognostic factors that affected OS were significantly associated with worse survival (p<0.05), namely, AFP level, macrovascular invasion, Edmondsone Steiner grade, surgical margin and Ishak score. AFP level, macrovascular invasion, microvascular invasion, and surgical margin influenced disease-free survival independently (p<0.05). Conclusions: The safety of hepatectomy for huge HCC is similar to that for large and small HCC; and this approach for huge HCC may achieve similar long-term survival and disease-free survival as for large HCC.