• Journal of Preventive Medicine and Public Health
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• v.43 no.2
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• pp.99-104
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• 2010

The World Health Organization (WHO) announced the emergence of a novel influenza on April 24, 2009, and they declared pandemic on June 11. In Korea, the proportion of influenza-like illness and the consumption of antiviral agents peaked in early November. The government established the Central Headquarters for Influenza Control and operated the emergency response system. In the quarantine stations, we checked the body temperature and collected quarantine questionnaires from all the arrivals from infected countries. We also isolated the confirmed cases in the national isolation hospitals. However, as the community outbreaks were reported, we changed strategy from containment to mitigation. We changed the antiviral agent prescription guideline so that doctors could prescribe antiviral agents to all patients with acute febrile respiratory illness, without a laboratory diagnosis. Also the 470 designated hospitals were activated to enhance the efficacy of treatment. We vaccinated about 12 million people and manage the adverse event following the immunization management system. In 2010, we will establish additional national isolation wards and support hospitals to establish fever clinics and isolation intensive care unit (ICU) beds. We will also make a computer program for managing the national isolation hospitals and designated hospitals. We will establish isolation rooms and expand the laboratory in quarantine stations and we will construct a bio-safety level 3 laboratory in each province. In addition, we plan to construct a bio-safety level 4 laboratory at a new Korea Centers for Disease Control and Prevention (KCDC) facilities in Ossong.

• Journal of Korean Society of Steel Construction
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• v.13 no.3
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• pp.313-325
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• 2001

Tubular bridge rail was developed to restrain and redirect a 14ton van-type truck. The developed bridge rail permits better visibility than concrete safety-shape bridge rail, and it has better structural adequacy than the existing steel and aluminum bridge rails in Korea. The new bridge rail consists of a tubular thrie beam(TTB) rail and a steel guard rail, which are connected to composite posts. The TTB shape provides both better containment of diverse bumper heights and more tight fit between the ends of bridge rail and roadside guardrails than the existing bridge rail sections currently used in Korea. Making composite post by filling concrete inside the steel pipe of the same size as are used for the roadside guardrail post was found to be more efficient in increasing the stiffness and ultimate strength than simply increasing the size of the steel pipe. The system was crash-tested for the impact condition of 14ton-80km/h-$15^{\circ}$, and it satisfied all evaluation criteria set forth in NCHRP Report 350 for a Test Level 4 safety appurtenance. Acceptable performances were obtained in computer simulations for the impact condition of S2.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.355-361
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• 2014

Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.3
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• pp.136-142
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• 2019

Nuclear power plant concrete structures are in contact with the coast, and durability due to chloride attack is very important because it is used as cooling water by taking seawater. For this purpose, a 3-year long-term saltwater immersion test was carried out to evaluate chloride ion diffusion coefficient and age apponent (m) The m values of the foundation with 4,000 class was 0.35 ~ 0.39, similar to KCI or ACI suggested values. essential service water constructions and tunnels of 5,000 class were 0.44 ~ 0.53 and 6,000 class, and 0.62 of reactor containment buildings were similar to the proposed values of FIB. As a result of the prediction of the service life with the measured age coefficient, all the safety related concrete structures of the nuclear power plants satisfied the service life of more than 60 years.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.2
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• pp.157-178
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• 2013

This study gives a conceptual and basic direction to develop a URL (underground research laboratory) program for establishing the performance and safety of a deep geological disposal system in Korea. The concept of deep geological disposal is one of the preferred methodologies for the final disposal of spent nuclear fuel (SNF). Advanced countries with radioactive waste disposal have developed their own disposal concepts reasonable to their social and environmental conditions and applied to their commercial projects. Deep geological disposal system is a multi-barrier system generally consisting of an engineered barrier and natural barrier. A disposal facility and its host environment can be relied on a necessary containment and isolation over timescales envisaged as several to tens of thousands of years. A disposal system is not allowed in the commercial stage of the disposal program without a validation and demonstration of the performance and safety of the system. All issues confirming performance and safety of a disposal system include investigation, analysis, assessment, design, construction, operation and closure from planning to closure of the deep geological repository. Advanced countries perform RD&D (research, development & demonstration) programs to validate the performance and safety of a disposal system using a URL facility located at the preferred rock area within their own territories. The results and processes from the URL program contribute to construct technical criteria and guidelines for site selection as well as suitability and safety assessment of the final disposal site. Furthermore, the URL program also plays a decisive role in promoting scientific understanding of the deep geological disposal system for stakeholders, such as the public, regulator, and experts.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.56-68
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• 2016

Extensive researches have been carried out for enhancing the safety of nuclear power plants and, especially, the development of passive cooling systems, such as passive containment cooling system (PCCS) and passive residual heat removal system, is increasingly important, where condensation is a crucial heat transfer mechanism. Recently, Ahn & Yun et al. developed a horizontal in-tube condensation heat transfer model as one of the activities for the PCCS development. In this work, we implemented the Ahn & Yun 's condensation heat transfer model into the MARS code and assessed it using the PASCAL experimental data. Based on the results of the assessment, we identified the limitations of the Ahn & Yun 's model and suggested a modified Ahn & Yun 's model, and assessed the model using various experimental data.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.1 s.173
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Journal of Energy Engineering
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• v.11 no.3
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• pp.237-246
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• 2002

If the safety depressurization system of APR-1400, the Korean next generation reactor, is in operation, water, air and steam are successively discharging into a in-containment refueling water storage tank through spargers. Among the phenomena occurring during the discharging processes, the air bubble clouds produce a low-frequency and high-amplitude oscillatory loading, which may result in the most significant damages to the submerged structures if the oscillation frequency is the same or close to the natural frequency of the structures. The involved phenomena are so complicated that most of the prediction of frequency and pressure loads has been resorted to experimental work and computational approach has been precluded. This study deals with a numerical simulation on the behavior of air bubble clouds discharging into a water pool through a sparger, by using a commercial thermal hydraulic analysis code, FLUENT, version 4.5. Among the multiphase flow models, the VOF (Volume Of Fluid) model was selected to simulate the water, air and steam flows. A satisfactory result was obtained comparing the analysis results with the ABB-Atom test results which had been performed for the development of sparser.