• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.175-182
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• 2010

As part of a fundamental study on the volume reduction of contaminated concrete wastes, the separation characteristics of the aggregates and the distribution of the radioactivity in the aggregates were investigated. Radioisotope $^{60}Co$ was artificially used as a model contaminant for non-radioactive crushed concrete waste. Volume reduction for radioactively contaminated dismantled concrete wastes was carried out using activated heavy weight concrete taken from the Korea Research Reactor 2 (KRR-2) and light weight concrete from the Uranium Conversion Plant (UCP). The results showed that most of the $^{60}Co$ nuclide was easily separated from the contaminated dismantled concrete waste and was concentrated mainly in the porous fine cement paste. The heating temperature was found to be one of the effective parameters in the removal of the radionuclide from concrete waste. The volume reduction rate achieved was above 80% for the KRR-2 concrete wastes and above 75% for the UCP concrete wastes by thermal and mechanical treatment.

• Journal of the Korea institute for structural maintenance and inspection
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• v.26 no.5
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• pp.112-118
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• 2022

In this study, the diffusion characteristics were evaluated using the concrete mix design of nuclear safety-related structures. Among the concrete structures related to nuclear power safety, we selected the composition of intake and drainage structures that are immersed in seawater or located on the tidal platform and evaluated the chloride ion permeation resistance by compressive strength and electrical conductivity and the diffusion characteristics by immersion in salt water. analyzed. Compressive strength was measured on the 1st, 7th, 14th, 28th, 56th, and 91st days until the 91st day, which is the design standard strength of the nuclear power plant concrete structure, and chloride ion permeation resistance was evaluated on the 28th and 91st. After immersing the 28-day concrete specimens in salt water for 28 days, the diffusion coefficient was derived by collecting samples at different depths and analyzing the amount of chloride. As a result, it was found that after 28 days, the long-term strength enhancement effect of the nuclear power plant concrete mix with 20% fly ash replacement was higher than that of concrete using 100% ordinary Portland cement. It was also found that the nuclear power plant concrete mix has higher chloride ion permeation resistance, lower diffusion coefficient, and higher resistance to salt damage than the concrete mix using 100% ordinary Portland cement.

• Journal of the Korea institute for structural maintenance and inspection
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• v.27 no.5
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• pp.16-22
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• 2023

In this study, a salt water immersion test was performed on concrete specimens simulating the concrete mix design of the nuclear power plant, and the correlation between the amount of chloride and the XRF component according to the depth of the concrete was analyzed. The amount of chloride on the surface of the nuclear power plant concrete increased slightly with increasing immersion time in salt water, but the amount of chloride in the depth of 5.5 mm or more showed a clear tendency to increase with increasing immersion time in salt water. As a result of analyzing the correlation between the amount of chloride in concrete and the XRF component, the concrete with 20% FA substitution compared with the OPC concrete showed a very high correlation between the composition ratio of Cl ions and the evaluation result of salt damage resistance by XRF component analysis. Accordingly, it was confirmed that chlorine ion analysis and salt damage resistance performance evaluation by XRF component analysis were possible through repeated data accumulation in the nuclear power plant concrete mix with 20% fly ash replacement.

• Computers and Concrete
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• v.12 no.5
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• pp.627-650
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• 2013

This paper deals with application of a non-linear continuum model for reinforced concrete affected by alkali-aggregate reaction (AAR) to analysis of some nuclear structures. The macroscopic behaviour of the material affected by AAR is described by incorporating a homogenization/averaging procedure. The formulation addresses the main stages of the deformation process, i.e., a homogeneous deformation mode as well as that involving localized deformation, associated with formation of macrocracks. The formulation is applied to examine the mechanical behaviour of some reinforced concrete structures in nuclear power facilities located in Quebec (Canada). First, a containment structure is analyzed subjected to 45 years of continuing AAR. Later, an inelastic analysis is carried out for the spent fuel pool taking into account the interaction with the adjacent jointed rock mass foundation. In the latter case, the structure is said to be subjected to continuing AAR that is followed by a seismic event.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.151-158
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• 2005

For the safety analysis of large structures such as nuclear containment buildings, we conventionally prefer to use analytical approach using finite element method rather than empirical test. Therefor, this paper is mainly focused to develop low-order solid finite element model with the elasto-plastic material model for the safety analysis of nuclear containment building. Drucker-Prager failure criteria in uncracked concrete and maximum tensile stress criteria in cracked concrete are used to model the constitutive behavior of concrete. The concrete material model takes into account the aspects of tensile strain, compression strength reduction of concrete and shear transfer to improve the accuracy of the finite element analysis. Finally, numerical simulation to compare the performance of the developed model with experimental results is employed. The numerical results in this study agree very well with the experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.669-674
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• 1996

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2019-2024
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• 2021

For evaluating the radiological safety of dismantled concrete, the process of disposal and recycling of the radioactive concrete generated during the dismantling of Kori Unit 1 is analyzed. Four scenarios are derived based on the analysis of the concrete recycling and disposal process, and the potential exposure to the workers and public during this process are calculated. VISIPLAN and RESRAD code are used for evaluating the dosages received by the workers and public in the following four scenarios: concrete inspection, transport of concrete by the truck driver, driving on a recycled concrete road, and public living near the landfilled concrete waste. Two worker exposure scenarios in the processing of concrete and two public exposure scenarios in recycling and disposal are considered; in all the scenarios, the exposure dose does not exceed the annual dose limit for each representative.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.336-347
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• 1997

During postulated severe accidents in Light water Reactors, molten corium which was ejected from the reactor vessel bottom, may erode the concrete basemat of the containment and there by threaten the containment integrity. This study experimentally examines the molten core-concrete interaction (MCC) using 20kg of thermite melt (Fe + $Al_2$O$_3$) and the concrete, used in Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 & 4) in Korea. The measured data are the downward heat fluxes, concrete erosion rate, gases and particle generation rates during MCCI. Transient results ore compared with those of TURCIT experiment conducted by SNL in USA. The peak downward heat flux to the concrete was measured to be about 2.1㎿/$m^2$. The initial concrete erosion rate was 175cm per hour, decreasing to 30cm per hour. It was shown from the post-test that the erosion was progressed downward up to 18mm in the concrete slug.

• International conference on construction engineering and project management
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• 2015.10a
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• pp.600-601
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• 2015

Both the importance and process of estimating Nuclear-power plant construction time and cost have increased in significance as energy user costs themselves have become more significant. In estimating construction time, few parameters are more significant than work item production rates and factors significantly affecting the rates. A standardized data collection tool was used to acquire a total of 401 data points from a S Nuclear-power plant project, for selected critical works: form-work, rebar-work, and concrete-pouring. With the data, several hypothesized drivers of the man-hour production rates and crew-day production rates were also analyzed. Findings from this study will enable industry professionals to enhance accuracy of time and cost estimation for nuclear power plant construction.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2023.05a
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• pp.315-316
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• 2023

In the present investigation, the chloride ion penetration resistance of nuclear power plant concrete with varying water-to-cement ratios was assessed. A comparative analysis was conducted on concretes that do not incorporate supplementary cementitious materials, such as fly ash, using permanently decommissioned nuclear structures as a reference. The objective is to employ this acquired data as a fundamental resource for the evaluation of the residual service life of nuclear power plant structures in subsequent studies.