• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.1053-1064
• /
• 2009

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.25 no.2
• /
• pp.53-58
• /
• 2021

Nuclear power plant's safety against seismic events is evaluated as risk values by probabilistic seismic safety assessment. The risk values vary by the seismic failure correlation between the structures, systems, and components (SSCs). However, most probabilistic seismic safety assessments idealized the seismic failure correlation between the SSCs as entirely dependent or independent. Such a consideration results in an inaccurate assessment result not reflecting real physical phenomenon. A nuclear power plant's seismic risk should be calculated with the appropriate seismic failure correlation coefficient between the SSCs for a reasonable outcome. An accident scenario that has an enormous impact on a nuclear power plant's seismic risk was selected. Moreover, the probabilistic seismic response analyses of a nuclear power plant were performed to derive appropriate seismic failure correlations between SSCs. Based on the analysis results, the seismic failure correlation coefficient between SSCs was derived, and the seismic fragility curve and core damage frequency of the loss of essential power event were calculated. Results were compared with the seismic fragility and core damage frequency of assuming the seismic failure correlations between SSCs were independent and entirely dependent.

• Journal of Ocean Engineering and Technology
• /
• v.32 no.5
• /
• pp.310-323
• /
• 2018

In order to protect lives and prevent large-scale injuries in the event of a fire on a ship or an offshore plant, most classification societies are strengthening their fire resistance designs of relevant cargo holds and accommodation compartments to keep flames from being transferred from a fire point to other compartments. Particularly in critical compartments, where flames should not propagate for a certain period of time, such as the A60 class division, both the airtightness and fire-resistant design of a piece passing through a bulkhead are subject to the Safety of Life at Sea Convention (SOLAS) issued by the International Maritime Organization (IMO). In order to verify the suitability of a fire-resistant design for such a penetrating piece, the fire test procedure prescribed by the Maritime Safety Committee (MSC) must be carried out. However, a numerical simulation should first be conducted to minimize the time and cost of the fire resistance test. In this study, transient thermal analyses based on the finite element method were applied to investigate the heat transfer characteristics of a bulkhead penetration piece for the A60 class compartment. In order to determine a rational bulkhead penetration piece design, the transient heat transfer characteristics according to the variation of design parameters such as the diameter, length, and material were reviewed. The verification of the design specification based on a numerical analysis of the transient heat transfer performed in this study will be discussed in the following research paper for the actual fire protection test of the A60 class bulkhead penetration piece.

• Journal of the Korea Institute of Information and Communication Engineering
• /
• v.25 no.10
• /
• pp.1381-1390
• /
• 2021

Smart grid is a representative convergence new technology in the era of the 4th industry revolution that deals with three broad security areas consisting of control system, the power grid, and the consumer. As it is a convergence new technology of the 4th industrial society, it is true that it can have a positive effect on the country's technological development, growth engine, and economic feasibility in the future. However, since the smart grid is expected to cause enormous damage in the event of a security accident, energy-related organizations must prepare various security measures to predict and respond to the latest security incidents. In this paper, the current status of domestic and foreign smart grids, trends in security standards, vulnerabilities and threats, and prospects for smart grid security technologies are to be considered.

• Nuclear Engineering and Technology
• /
• v.56 no.1
• /
• pp.132-140
• /
• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.64 no.12
• /
• pp.1768-1775
• /
• 2015

Recent major domestic facilities, such as nuclear power plants, many control cables are installed and are degraded by long-term use, but research on deterioration diagnosis is lacking. In the event of a fault in the cable due to deterioration can be developed into a major accident such as the main plant is stopped, so the deterioration diagnostic techniques of high reliability for the cable is required. In this paper, proposes a methodology using a cable resonance that can effectively diagnose the deterioration of the cable. Prior to the test, we developed a setup for stable measuring the characteristics of the cable and it verified the suitable of the measurement set-up in terms of interactivity and reliability, also measured S-parameters applying verified measurement set-up to the cables that deterioration degree is different. Then, we had amplified the difference in resonance frequency between the healthy state and the deteriorated state using connection in a series of measured S-parameters. In a result from the method, we have verified that the more deteriorate the cables is, the more decrease the resonance frequency is. Measured results are justified by inducing the resonance frequency calculation of the cable from the S- parameters represented by the hyperbolic function formula. VNA(Vector Network Analyzer) for S-parameter measurements used in this study is Agilent E5061B and shielded twisted-pair cables was used for deterioration diagnostic test.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
• /
• 2015.10a
• /
• pp.1013-1014
• /
• 2015

Regardless of the strict enforcement of prevention education in accordance with the school facility management standard, safety accidents that lead to human and physical damages occur in current educational facilities because of teenagers with very low sense of responsibility and insensitivity toward fire and facility safety. To ensure educational facility safety, technology that will enable a fast work process and easy confirmation of electronic blueprints and related documents about the educational facility through smart devices at the site by various means is needed. This paper proposes a system design linked to the National Education Information System (NEIS) that uses the document conversion function, high efficiency resolution, and Internet of Things (IoT) to inspect and control the educational facility in the event of a safety accident through the Educational Facility Resource Management System (EFRMS) that manages the electronic blueprints, and various educational facility documents through various smart devices.

• Journal of the Korean Society of Safety
• /
• v.35 no.4
• /
• pp.74-83
• /
• 2020

In Korea, about 110,000 LPG small storage tanks of less than three tons have been installed in restaurants, houses and factories, and are used as LPG supply facilities for cooking, heating and industrial use. In the case of combustible liquefied gas storage tanks, the tank may rupture due to the temperature increase of the tank steel plate (approximately 600℃) even when the safety valve is operating normally, causing large-scale damage in an instant. Therefore, in the event of a fire near the LPG small storage tank, it is necessary to accurately predict the timing of the BLEVE(Boiling Liquid Expanding Vapour Explosion) outbreak in order to secure golden time for lifesaving and safely carry out fire extinguishing activities. In this study, we have first investigated the results of a prior study on the prediction of the occurrence of BLEVE in the horizontal tanks. And we have developed thermodynamic models and simulation program on the prediction of BLEVE that can be applied to vertical tanks used in Korea, have studied the effects of the safety valve's ability to vent, heat flux strength of external fires, size of tanks, and gas remaining in tanks on the time of BLEVE occurrence and have suggested future utilization measures.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2077-2083
• /
• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.

• Journal of the Korea Institute of Information and Communication Engineering
• /
• v.18 no.7
• /
• pp.1671-1680
• /
• 2014

There are always much concern about the leakage of radiation materials in the event of dismantle or unexpected accident of nuclear power plant. In order to remove the leakage of radiation materials, appropriate dispersion detection techniques for radiation materials are necessary. However, because direct handling of radiation materials is highly restricted and risky, developing radiation-related techniques needs computer simulation in advance to evaluate the feasibility. In this paper, we propose a radiation imaging technique which can acquire 3D dispersion information of radiation materials and tested by simulation. Using two virtual 1D radiation sensors, we obtain stereo radiation images and acquire the 3D depth to virtual radiation materials using stereo disparity. For point and plane type virtual radiation materials, the possibility of the acquisition of stereo radiation image and 3D information are simulated.