• Nuclear Engineering and Technology
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• 제55권2호
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• pp.530-545
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• 2023

As an alternative to conventional management options for a lot of concrete waste from decommissioning of nuclear power plants, a set of scenarios for controlled recycling of decommissioning concrete waste as engineered barriers of a radioactive waste repository was proposed, and a comprehensive safety assessment model and framework covering both pre-and post-closure phases was newly developed. The new methodology was applied to a reference vault-type repository, and the ratios of derived concentration limits to unconditional clearance levels of eighteen radionuclides for controlled recycling were provided for three sets of dose criteria (0.01, 1, and 20 mSv/y for the pre-closure and 0.01 mSv/y for the post-closure phases). It turns out that decommissioning concrete waste whose concentration is much higher than the unconditional clearance level can be recycled even when the dose criterion 0.01 mSv/y is applied. Moreover, a case study on ABWR bio-shield shows that the fraction of recyclable concrete waste increases significantly by increasing the dose criterion for the radiation worker in the pre-closure phase or the duration of storage prior to recycling. The results of this study are expected to contribute to demonstrating the feasibility of controlled recycling of a lot of decommissioning concrete waste within nuclear sectors.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2079-2091
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• 2024

In this paper, a model is developed for calculating pH in the presence of organic impurities due to dissolution of paint and/or continuous injection of organic impurities in the sump. The model is implemented in the AnCheBi code for the analysis of chemical behaviors of the iodine in the containment when the pH changes during a severe accident. Validation of the model is performed with P10T2 and P11T1 experiments carried out by AECL in Canada under the BIP project. Importance analyses of the pH calculation model in the AnCheBi code are then performed with the aforementioned experimental data via Latin hypercube sampling on the reaction coefficients, sensitivity analyses of AnCheBi, and calculation of the correlation coefficients between the reaction coefficients and figure of merits (the pH and the concentrations of the various iodine species). From the importance analyses, we provide the sensitivity of the pH calculation model to the change of pH and the concentrations of the various iodine species and the reaction coefficients related with the dominant phenomena underlying the change of pH and the concentrations of the species.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1321-1329
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• 2016

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from $450^{\circ}C$ to $700^{\circ}C$ and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3310-3316
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• 2022

Owing to the rising concerns about the safety of nuclear power plants (NPP), it is essential to study the venturi scrubber in detail, which is a key component of the filtered containment venting system (FCVS). FCVS alleviates the pressurein containment byfiltering and venting out the contaminated air. Themain objective of this research was to perform a CFD investigation of different configurations of a circular, non-submerged, self-priming venturi scrubber to estimate and improve the performance in the removal of elemental iodine from the air. For benchmarking, a mass transfer model which is based on two-film theory was selected and validated by experimental data where an alkaline solution was considered as the scrubbing solution. This mass transfer model was modified and implemented on a unique formation of two self-priming venturi scrubbers in series. Euler-Euler method was used for two-phase modeling and the realizable K-ε model was used for capturing the turbulence. The obtained results showed a remarkable improvement in the removal of radioactive iodine from the air using a series combination of venturi scrubbers. The removal efficiency was improved at every single data point.

• 에너지공학
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• 제22권1호
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• pp.8-16
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• 2013

A core catcher is considered as a promising engineered system to stabilize the molten corium in the containment during a postulated severe accident in a nuclear power plant. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in ex-vessel core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single-phase natural circulation in the whole core catcher system. In this study, the analysis of the test facility for the core catcher using the CUPID code, which is a three-dimensional thermal-hydraulic code for the simulation of two-phase flows, was carried out to evaluate its cooling capacity.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.75-81
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• 1996

COBRA-TF is a two fluid, three field subchannel code. Three fields are continuous vapor, continuous liquid and droplet. Some assessments are conducted to validate the related models and to estimate a code ability through dryout and post-CHF experiment in a tube and DNB test in rod bundles. It turned out form dryout and post-CHF experiment that the predicted dryout locations and wall temperature profiles are in close agreement with the experiments. On the other hand, DNB prediction of COBRA-TF are performed for two kinds of rod bundles along with EPRI CHF correlation. To estimate its performance COBRA-IV of homogeneous model is also run for the same data. The results say that COBRA-TF/EPRI is better in DNB prediction than COBRA-IV/EPRI. In addition the thermal-hydraulic behaviors due to the different two-phase flow models are presented at the condition of CHF.

• Nuclear Engineering and Technology
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• 제33권3호
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• pp.261-269
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• 2001

In this study, sorption experiments of uranium onto silica colloids were carried out and the effects of important geochemical parameters such as pH, ionic strength, carbonate concentration, colloid concentration, and total concentration of uranium were investigated. The sorption coefficients of uranium for silica colloids named as pseudo-colloid formation constants were about 10$^4$~ 10$^{5}$ mL/g depending on the experimental conditions. The effects of the geochemical parameters were found to be important in the sorption of uranium onto silica colloids. A Langmuir type sorption isotherm of uranium between silica colloids and the solution phase was also presented. The sorption mechanisms were explained by analyzing the effects of the geochemical parameters.

• 한국소음진동공학회논문집
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• 제19권4호
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• pp.370-377
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• 2009

A simple preliminary analysis is often useful to check a validity of design alternatives before the detailed analysis phase in the viewpoint of efficiency. This paper describes a preliminary analysis procedure for the selection among basket design candidates for the spent fuel shipping cask of Korean standard nuclear power plant. As the cask should maintain the structural integrity in hypothetical accident condition, the case of 9 m drop is significantly considered as the worst scenario among the accident conditions in structural design viewpoint in this paper. As basket design options, totally four different types are considered and analyzed in the point of structural integrity at drop impact and weldability for fabrication. As a result, an insertion round plate type with densely spaced supports turns out to be the best in both of the viewpoints, though the weld plate type shows a bit more design margin.

• 한국태양에너지학회 논문집
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• 제22권1호
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• pp.43-53
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• 2002

This study has been carried out to study the thermal performance of an all-glass type solar collector tube when a heat transfer medium is used with a heat storage unit capable of preventing pressure build-up within the tube. The heat storage unit is devised such that it performs the dual function of relieving excessive pressure and storing solar thermal energy. Different types of heat storage medium are tested under heat-up phase of a collector tube. It is found that the proposed unit could be used quite effectively if one wishes to capitalize more aggressively in harnessing the solar energy.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.587-596
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• 2012

The present paper deals with the participation of the University of Pisa in the last International Standard Problem (ISP) focused on system thermal hydraulic, which was led by the Korean Atomic Energy Research Institution (KAERI). The selected test was a Direct Vessel Injection (DVI) line break carried out at the ATLAS facility. University of Pisa participated, together with other eighteen institutions, in both blind and open phase of the analytical exercise pursuing its methodology for developing and qualifying a nodalization. Qualitative and quantitative analysis of the code results have been performed for both ISP-50 phases, the latter adopting the Fast Fourier Transfer Based Method (FFTBM). The experiment has been characterized by three-dimensional behavior in downcomer and core region. Even though an attempt to reproduce these phenomena, by developing a fictitious three-dimensional nodalization has been realized, the obtained results were generally acceptable but not fully satisfactory in replicating 3D behavior.