• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.462-463
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• 2018

• Journal of Institute of Control, Robotics and Systems
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• v.11 no.9
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• pp.774-780
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• 2005

For the accurate inspection of the resources and space in the ocean, the method of locating the measurement equipments in the seafloor and retrieving these equipments later after a certain period of time. is generally used. In this method, the reliability of retrieving measurement equipments is very important. In our proposed remotely-controlled acoustic release system, an underwater ultrasonic wave recognition algorithm by which we can recognize the sound signal without the influence of disturbances due to underwater environment changes is developed, and a battery is used for the reduction of electric power consumption. we show the effectiveness of our proposed system through experimental results.

• Fire Science and Engineering
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• v.16 no.2
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• pp.75-80
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• 2002

In this paper, some experiments of a heat release rate and toxicity for underground utility 22.9kv cable in fire was conducted and analysed applying plume equation and smoke chamber test separately, A 22.9 ㎸ power cable is selected for testing heat release in ISO 9705 geometry and toxicity production is measured with NES 713 (British-Naval Engineering Standard)test. In test results, Cable heat release reached about 60 ㎾ above 1.2 m from heptane pan and CO generated lethal concentration under 30 min. exposure condition.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.651-656
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• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• Proceedings of the Korean Institute of Navigation and Port Research Conference
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• 2014.10a
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• pp.87-88
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• 2014

The Coastal sediment is highly contaminated due to ship transportation, industries discharges and urban sources. Various contaminants release into seawater and settle in marine sediment and it significantly affect marine eco system. In the present study evaluated the optimization of slow release biostimulant ball (BSB) in coastal sediment in busan. The effective variables like BSB size, distance and month variables on VS reduction was determined by using Response surface methodology(RSM). The analysis of variance (ANOVA) and coefficient determination (R2) of VS was 0.9369 and maximum reduction of VS was obtained in 3cm ball size and 5.5cm distance and 4 month interval time. This result revealed that the BSB in effective VS reduction in coastal sediment.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1250-1258
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• 2021

In this paper, the mechanical performance of cracked surface-coated Zircaloy cladding, which has different coating materials, coating thicknesses and initial crack lengths, has been investigated. By analyzing the stress field near the crack tip, the safety zone range of initial crack length has been decided. In order to determine whether the crack can propagate along the radial (r) or axial (z) directions, the energy release rate has been calculated. By comparing the energy release rate with fracture toughness of materials, we can divide the initial crack lengths into three zones: safety zone, discussion zone and danger zone. The results show that Cr is suitable coating material for the cladding with a thin coating while Fe-Cr-Al have a better fracture mechanical performance in the cladding with thick coating. The Si-coated and SiC-coated claddings are suitable for reactors with low power fuel elements. Conclusions in this paper can provide reference and guidance for the cladding design of nuclear fuel elements.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.116-127
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• 1990

The fission gas release model used In the SPEAR-BETA fuel performance code was modified by use of effective thermal conductivity for cracked fuel and by laking Into account axial fission-gas mixing between the fuel-clad gap and the plenum. With use of this modified model the fission gas release was analyzed under various power ramping conditions of P$_{max}$ and $\Delta$.fP. Effective fuel thermal conductivity that accounts for the effect of fuel tracking was used in calculation of the fuel temperature distribution and the Internal gas pressure under power ramping conditions. Mixing and dilution effects due to axial gas flow were also considered in computing the width and the thermal conductivity of the gap. The effect of axial gas flow w3s solved by the Crank-Nicholson method. The finite difference method was used to save running time in the calculation. The present modified fission-gas release model was validated by comparing its predicted results with experimental data from various lamping tests In the literature and calculated results with use of the models used In the SPEAR-BETA and FEMAXI-IV codes. Results obtained with use of the present modified model showed better agreement with experimental data reported in the literature than those results with use of the latter codes. The fuel centerline temperature calculated with introduction of effective thermal conductivity for centerline temperature calculated with Introduction of effective thermal conductivity for cracked fuel was 200 higher fission gas release predicted with use of the modified model was nearly 6% larger on the average than that calculated by use of the unmodified model used in the SPEAR-BETA code.e SPEAR-BETA code.e.

• Journal of Soil and Groundwater Environment
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• v.23 no.4
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• pp.1-15
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• 2018

Despite of safety issues related to radiological hazards, 31 countries around the world are operating more than 450 nuclear power plants (NPPs). To operate NPPs safely, safety regulations from radiation protection organizations were developed and adopted in many countries. However, many cases of radionuclide releases at foreign NPPs have been reported. Almost all commercial NPPs routinely release radioactive materials to the surrounding environments as liquid and gas phases under control. These releases are called 'planned releases' which are planned, regularly monitored, and well documented. Meanwhile, the releases focused in this review, called 'unplanned releases', are neither planned nor monitored by regulatory and/or protection organizations. NPPs are generally composed of various structures, systems and components (SSCs) for safety. Among them, the SSCs near reactors are closely related to safety of NPPs, and typically fabricated to comply with stringent requirements. However, some non-safety related SSCs such as underground pipes may be constructed only according to commercial standards, causing the leakage of radioactive fluids usually containing tritium ($^3H$). This paper discusses SSCs of NPPs and introduces several cases of unplanned releases at foreign NPPs. The current regulation on the environmental radiological surveillance and assessment around the NPPs in South Korea are also examined.

• Journal of the Korean Society of Radiology
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• v.11 no.1
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• pp.1-8
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• 2017

The U.S. NRC establishes a radiological criteria with regard to restricted or unrestricted use of nuclear plant site after decommissioning in NUREG-1757. According to this, a nuclear plant site can be released in a restricted way or unrestricted way only if a licensee demonstrates that the dose criteria is fulfilled after the site decontamination and remediation. In order to prove compliance with the radiological criteria of site release, LTP(License Termination Plan) must include the site release criteria, site characterization, final survey plan with major radionuclides and DCGL(Derived Concentration Guideline Levels), etc. Based on the decommissioning case of Rancho Seco nuclear power plant in the United States, this paper analyzed a method of setting the DCGL that can be applied to Kori NPP Unit 1 which will be permanently disabled in 2017.

• Fire Science and Engineering
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• v.20 no.2 s.62
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• pp.14-20
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• 2006

The aim of this study is to predict the ignition times and the HRR(heat release rate) for building interior materials. By using the literature data and RSM(response surface methodology), the new equations for predicting the ignition time and the HRR of building interior materials are proposed. The A.A.P.E.(average absolute percent error) and the A.A.D.(average absolute deviation) of the reported and the calculated ignition times by means of the thickness and the density were 4.35 sec and 1.57 sec, and the correlation coefficient was 0.987. The correlation coefficient of the reported and the calculated the net HRR by means of burner width and power was 0.983. Also the correlation coefficient of the reported and the calculated the total HHR by means of burner width and power was 0.999. The correlation coefficient of the reported and the calculated the maximum flame height by means of burner width and power was 0.999. The values calculated by the proposed equations were in good agreement with the literature data.