• Journal of the Korea Safety Management & Science
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• v.9 no.6
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• pp.123-135
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• 2007

In the recent years, the companies have manually recorded a production status in a work diary or have mainly used a bar code in order to collect each process's progress status, production performance and quality information in the production and logistics process in real time. But, it requires an additional work because the worker's record must be daily checked or the worker must read it with the bar code scanner. At this time, data's accuracy is decreased owing to the worker's intention or mistake, and it causes the problem of the system's reliability. Accordingly, in order to solve such problem, the companies have introduced RFID which comes into the spotlight in the latest automatic identification field. In order to introduce the RFID technology, the process flow must be analyzed, but the ASME sign used by most manufacturing companies has the difficult problem when the aggregation event occurs. Hence, in this study, the RFID logistic flow analysis Modeling Notation was proposed as the signature which can analyze the manufacturing logistic flow amicably, and the manufacturing logistic flow by industry type was analyzed by using the proposed RFID logistic flow analysis signature. Also, to monitor real-time information through EPCglobal network, EPCISEvent template by industry was proposed, and it was utilized as the benchmarking case of companies for RFID introduction. This study suggested to ensure the decision-making on real-time information through EPCglobal network. This study is intended to suggest the Modeling Notation suitable for RFID characteristics, and the study is intended to establish the business step and to present the vocabulary.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.22 no.5
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• pp.554-560
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• 2021

CRUD (chalk river unidentified deposits) is a porous material deposited on the surface of nuclear fuel during nuclear power plant operation. The CRUD is composed of metal oxides, such as iron, nickel, and chromium. It is essential to investigate the effects of the CRUD layer on the wall heat transfer between the nuclear fuel surface and the coolant in the event of a nuclear accident. CRUD only negatively affects the temperature of the nuclear fuel due to heat resistance because the effects of the CRUD layer on two-phase boiling heat transfer are not considered. In this study, the physical property models for the porous CRUD layer were developed and implemented into the SPACE code. The effects of boiling heat transfer models on the peak cladding temperature and quenching were investigated by simulating a reflood experiment. The calculation results showed some positive effects of the CRUD layer.

• Journal of the Korean Society of Safety
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• v.36 no.3
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• pp.74-80
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• 2021

Korea Atomic Energy Research Institute (KAERI) has been operating an integral effects test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), according to APR1400 for transient experimental and design basis accident simulation. Moreover, based on the experimental data, the domestic standard problem (DSP) program has been conducted in Korea to validate system codes. Recently, through DSP-05, the performance of the passive auxiliary feedwater system (PAFS) in the event of multiple steam generator tube rupture (MSGTR) has been analyzed. However, some errors exist in the reference input model distributed for DSP-05. Furthermore, the calculation results of the heat loss correlation for the secondary system presented in the technical report of the reference indicate that a large difference is present in heat loss from the target value. Thus, in this study, the reference model is corrected using the geometric information from the design report and drawings of ATLAS. Additionally, a new heat loss correlation is suggested by fitting the results of the heat loss tests. Herein, MSGTR-PAFS accident analysis is performed using MARS-KS 1.5 with the improved model. The steady-state calculation results do not significantly differ from the experimental values, and the overall physical behavior of the transient state is properly predicted. Particularly, the predicted operating time of PAFS is similar to the experimental results obtained by the modified model. Furthermore, the operating time of PAFS varies according to the heat loss of the secondary system, and the sensitivity analysis results for the heat loss of the secondary system are presented.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Journal of Radiation Protection and Research
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• v.48 no.3
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• pp.131-143
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• 2023

Background: This study examined the detection limit of thyroid screening monitoring conducted at the time of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in 2011 using a Monte Carlo simulation. Materials and Methods: We calculated the detection limit of a NaI(Tl) survey meter to measure ¹³¹I accumulation in the thyroid gland of children. Mathematical phantoms of 1- and 5-year-old children were developed in the simulation of the Particle and Heavy Ion Transport code System code. Contamination of the body surface with eight radionuclides found after the FDNPP accident was assumed to have been deposited on the neck and shoulder area. Results and Discussion: The detection limit was calculated as a function of ambient dose rate. In the case of 40 Bq/cm² contamination on the body surface of the neck, the present simulations showed that residual thyroid radioactivity corresponding to thyroid dose of 100 mSv can be detected within 21 days after intake at the ambient dose rate of 0.2 µSv/hr and within 11 days in the case of 2.0 µSv/hr. When a time constant of 10 seconds was used at the dose rate of 0.2 µSv/hr, the estimated survey meter output error was 5%. Evaluation of the effect of individual differences in the location of the thyroid gland confirmed that the measured value would decrease by approximately 6% for a height difference of ±1 cm and increase by approximately 65% for a depth of 1 cm. Conclusion: In the event of a nuclear disaster, simple measurements carried out using a NaI(Tl) scintillation survey meter remain effective for assessing ¹³¹I intake. However, it should be noted that the presence of short-half-life radioactive materials on the body surface affects the detection limit.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.645-660
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• 1995

The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.14 no.2
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• pp.245-255
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• 1994

A two-dimensional Lagrangian finite-difference computer program is developed for the wave propagation analysis of impact phenomena. The numerical scheme is the standard method originally proposed by Von Neuman and Richtmyer, using artificial viscosity to smooth shock fronts. The material model used in the study is the standard hydrodynamic-elastic-plastic relations with Von-Mises yield criterion. A test configuration consisted of a target and a projectile were calculated to understand the response of a colliding event. However, the computer code is in plane strain, the calculations were intended for generating the qualitative features of the model behaviors. Nevertheless, the computational results were consistent with the experimental observations and provided a rational basis to interpret the modes of failures.

• Computers and Concrete
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• v.21 no.3
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• pp.299-310
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• 2018

When reinforced concrete structures are subjected to strong seismic forces, their beam-column connections are very susceptible to be damaged during the earthquake event. Consequently, structural designers try to fit an important quantity of steel reinforcement inside the connection, complicating its construction without a clear justification for this. The aim of this work is to evaluate -and demonstrate- numerically how the quantity and the array of the internal steel reinforcement influences on the nonlinear response of the RC beam-column connection. For this, two specimens (extracted from an experimental test of 12 RC beam-column connections reported in literature) were modeled in the Finite Element code FEAP considering different stirrup's arrays. The nonlinear response of the RC beam-column connection is evaluated taking into account the nonlinear thermodynamic behavior of each component: a damage model is used for concrete; a classical plasticity model is adopted for steel reinforcement; the steel-concrete bonding is considered perfect without degradation. At the end, the experimental responses obtained in the tests are compared to the numerical results, as well as the distribution of shear stresses and damage inside the concrete core of the beam-column connection, which are analyzed for a low and high state of confinement.

• Convergence Security Journal
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• v.16 no.2
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• pp.63-70
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• 2016

With the development of information and communication technologies like internet, the environment where people are able to access internet at any time and at any place has been established. As a result, cyber threats have been tried through various routes. Of cyber threats, DDoS is on the constant rise. For DDoS prediction modeling, this study drew a DDoS security index prediction formula on the basis of event data by using a statistical technique, and quantified the drawn security index. It is expected that by using the proposed security index and coming up with a countermeasure against DDoS threats, it is possible to minimize damage and thereby the prediction model will become objective and efficient.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.60 no.6
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• pp.1255-1260
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• 2011

In this paper, we study on the reliability of Very High-rate Power Line Communication (VH-PLC) for Smart Grid, so that the resultant data rate is over 400Mbps at a physical layer. Firstly, reviewing the research trend of the PLC, we discuss the required techniques for supporting the Smart Grid. Considering a pre-specification with the value of several parameters, we investigate a multi-carrier modulation technique to overcome limitations of higher rate transmission under power line channel environments. Then, we propose a system specification of the VH-PLC in the sense of enhancing two features. One is resolving the problem of the co-existence of the deployed high-speed PLC according to the published standardization of KS X 4600-1 in Korea. The other is getting better performance on the grid adopting the diverse element techniques, such as multi-carrier modulation, a subcarrier utilization mode, a variable rate LDPC (Low Density Parity Check) code, and a time and frequency diversity technique. Further, a simulation tool, composed of an Event-Driven simulator and a Time-Driven simulator, is developed for the purpose of verifying the system performance and continuously cross-checking the test bench signal of the proposed VH-PLC system.