• Journal of the Korean Society for Advanced Composite Structures
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• v.6 no.2
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• pp.91-96
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• 2015

GRP pipe (Glass-fiber Reinforced Plastic Pipe) lines making use of FRP (Fiber Reinforced Plastic) are generally thinner, lighter, and stronger than the existing concrete or steel pipe lines, and it is excellent in stiffness/strength per unit weight. In this study, we present the result of field test for buried GRP pipes with large diameter(2,400mm). The vertical and horizontal ring deflections are measured for 387 days. The short-term deflection measured by the field test is compared with the result predicted by the Iowa formula. In addition, the long-term ring deflection is predicted by using the procedure suggested in ASTM D 5365(ANNEX) in the range of 40 to 60 years of service life of the pipe based on the experimental results. From the study, it was found that the long-term vertical and horizontal ring deflection up to 60 years is less than the 5% ring deflection limitation.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.15 no.4
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• pp.567-578
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• 2002

The main purpose of this study is to investigate the dynamic behaviour of containment building in nuclear power plant excited by aircraft impact loading using a lower order 8-node solid element. The yield and failure surfaces for concrete material model is formulated on the basis of Drucker-Prager yield criteria and are assumed to be varied by taking account of the visco-plastic energy dissipation. The standard 8-node solid element has prone to exhibit the element deficiencies and the so-called B bar method proposed by Hughes is therefore adopted in this study. The implicit Newmark method is adopted to ensure the numerical stability during the analysis. Finally, the effect of different levels of cracking strain and several types of aircraft loading are examined on the dynamic behaviour of containment building and the results are quantitatively summarized as a future benchmark.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.139-165
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• 2019

The current status and prospect of decommissioning technology development at KAERI are reviewed here. Specifically, this review focuses on four key technologies: decontamination, remote dismantling, decommissioning waste treatments, and site remediation. The decontamination technologies described are component decontamination and system decontamination. A cutting method and a remote handling method together with a decommissioning simulation are described as remote dismantling technologies. Although there are various types of radioactive waste generated by decommissioning activities, this review focuses on the major types of waste, such as metal waste, concrete waste, and soil waste together with certain special types, such as high-level and high-salt liquid waste, organic mixed waste, and uranium complex waste, which are known to be difficult to treat. Finally, in a site remediation technology review, a measurement and safety evaluation related to site reuse and a site remediation technique are described.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• Journal of the Korean Recycled Construction Resources Institute
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• v.10 no.3
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• pp.276-284
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• 2022

In this study, performance evaluation and rheological characteristics were analyzed for recycling the fine powder of nuclear power plant dismantled waste concrete as a solidifying agent for radioactive waste disposal. The radioactive concrete fine powder was used to prepare a simulated sample, and the test specimen was prepared using Di-water, CoCl₂, and 1 mol CsCl aqueous solution as mixing water. Regardless of the aggregate mixing ratio and the type of mixing water, it satisfies the performance standard of 3.45 MPa for compressive strength at 28 days of age. All specimens satisfied the criteria for submersion strength, and the thermal cycle compressive strength satisfies the criteria for all specimens except Plain-50. As a result of evaluating the rheological properties of the solidifying agent, it was found that the increase in the aggregate mixing rate decreased the yield stress and plastic viscosity. The leaching index for cobalt and cesium of all specimens was 6 or higher, which satisfies the standard. In order to secure the stable performance of the solidifying agent, it is considered effective to use 40 % or less of the aggregate component in the solidifying agent.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.12
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• pp.3748-3754
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• 2009

An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.11
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• pp.3260-3268
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• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

• Journal of Korean Society of Steel Construction
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• v.22 no.5
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• pp.487-496
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• 2010

The structural damping ratios for seismic design of nuclear power plant structures are specified in Regulatory guide 1.61 of the United States NRC for RC structures of 4%(OBE) and 7%(SSE), and for steel structures of 3%(OBE) and 4%(SSE), but not for steel-plate concrete (SC) structures that have been developed recently. The objective of this study is to investigate the damping ratios of SC structures by identifying the relative differences in the damping ratios between RC and SC structures. An experimental study was performed on four specimens, RC-S, RC-M, SC-S and SC-M, where S stands for shear-governed and M for moment-governed. The conducted method was free vibration testing by rupturing a brittle steel plate that linked the actuator and the mass center. The test results were analyzed to determine fundamental frequencies and damping ratios at various load levels. By examining the relative differences in damping ratios of four specimens, it is proposed for SC structures to use the same damping ratio of 4% as RC one at OBE, but 1% less damping ratio than RC one resulting in 6% at SSE.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.3
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• pp.303-309
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• 2005

Numerical analysis is carried out to identify the appropriateness of the design codes that is available for the tensile design of fastening system at Nuclear Power Plant (NPP) in this study. This study is intended for the cast-in-place anchor that is widely used for the fastening of equipment in Korean NPPs. The microplane model and the elastic-perfectly plastic model are employed for the quasi-brittle material like concrete and for the ductile material like anchor bolt as constitutive model for numerical analysis and smeared crack model is employed to simulate the clack and damage phenomena. The developed numerical model is verified on a basis of the various test data of cast-in-place anchor. The appropriateness of both ACI 349 Code and CEB-FIP Code is evaluated for the tensile design of cast-in-place anchor and it is proved that both design codes give a conservative results for real tensile capacity of cast-in-place anchor.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.30 no.6A
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• pp.543-560
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• 2010

Safety-related concrete structures in a nuclear power plant must be protected against the impact of flying objects, referred to in the profession as missiles. In practice, the structural verification is usually carried out by means of empirical formulas, which relate the velocity of the impinging missile to the wall thickness needed to prevent scabbing or perforation. The purpose of this study is to reevaluate the predictability of the local effect prediction formulas for the penetration and scabbing depths and perforation thickness. Therefore, available formulas for predicting the penetration depth, scabbing thickness, and perforation thickness of concrete structures impacted by solid missiles are summarized, reviewed, and compared. A series of impact analyses is performed to predict the local effects of the projectile at impact velocities varing from 95 to 215 m/s. The results obtained from the numerical simulations have been compared with tests that were carried out at Kojima to validate numerical modelling. The simulation results show reasonable agreement with the Kojima test results for the overall impact response of the RC slabs. From these results, it seems that the Degen equation give a very good estimate of perforation thickness against a tornado projectile for test data. Finally, the results obtained from the impact analysis have been compared with Degen formula to determine the perforation thickness of the RC slab.