• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.181-189
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• 2017

The purpose of this paper is to evaluate several leading options for the management of radioactive metallic waste against a set of general criteria including safety, cost effectiveness, radiological dose to workers and volume reduction. Several options for managing metallic waste generated from decommissioning are evaluated in this paper. These options include free release, controlled reuse, and direct disposal of radioactive metallic waste. Each of these options may involve treatment of the metal waste for volume reduction by physical cutting or melting. A multi-criteria decision analysis was performed using the Analytic Hierarchy Process (AHP) to rank the options. Melting radioactive metallic waste to produce metal ingots with controlled reuse or free release is found to be the most effective option.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.6
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• pp.930-937
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• 2003

The extension of design life among power plants is increasingly becoming a world-wide trend. Kori #1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts fur man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME Code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME Code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety fer helical-coil method which is used according to Code Case-N-496-1. From analysis results, we found .that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10％ by manufacture integrity criteria and 2／3 yield strength criteria on ASME Code. It was also confirmed that the helical-coil repair method would be safe.

• Water for future
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• v.23 no.4
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• pp.445-457
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• 1990

In order to predict the dispersion of a thermal discharge with strong turbulent and buoyant effects, the development of a numerical model using turbulence model and its application are significantly increased. In this study, a 3-dimensional steady-state model for the surface discharge of heated water into quiescent water body is developed. For the model closure of turbulent terms the 4-equation turbulence model is used. For economic numerical simulation, the elliptic governing equations are transformed to the partially parabolic equations. In general, the simulated results by the present model agree well to the experimental results by Pande and Rajaratnam. The model characteristics are presented in comparison with the predicted results of the 2-equation turbulence model by McGuirk and Rodi. Applying the 4-equation turbulence model to the Korea nuclear unit 1 at Kori site, feasibility and efficiency of the present model are validated.

• Computational Structural Engineering
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• v.10 no.4
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• pp.155-164
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• 1997

A reactor pressure vessel, which contains fuel assemblies and reactor vessel internals, has the thermal stress resulting from the cool-down and heat-up of the vessel wall in combination with the pressure stress from system pressure resulting in large stresses. The combination of the pressure stress and thermal stress along with a decrease in fracture toughness may cause through-wall propagation of a relatively small crack. Therefore, it is necessary to define the relations between operating pressure and temperature during cool-down and heat-up. In this study, theory of fracture mechanics for a pressure/temperature limit curve is investigated and a numerical procedure for generating it is developed. Plant-specific limit curves for the Kori unit 1 plant, the oldest nuclear power plant in Korea, have been obtained for several cooling and heating rates and their results are discussed.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.279-289
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• 1985

The purpose of this paper is to investigate the density-wave oscillation type instability in the recirculating loop of U-tube steam generator (UTSG). The perturbed and nodalized conservations equations based on the drift-flux model have been derived to obtain the single-and two-phase pressure drop perturbations, by taking into account the slip between phases, nonuniform heat flux and heated wall dynamics. To assess the stability, the frequency domain technique with the Nyquist criterion has been used under the constant pressure drop boundary condition through the loop. The computer implementation of this model, SASG, was used for the parametric study of the steam generator in Kori-Unit 1. The results of the parametric study revealed important factors influencing UTSG stability margin.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.243-253
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• 2022

Spent filters with a high radiation dose rate of 2 mSv·hr^-1 or more are not easily managed. So far, the Korean policy for spent filter disposal is to store them temporarily at nuclear power plants until the waste filters can be easily managed. Nuclear power plant decommissioning in Korea is starting with Kori unit 1. Volume reduction of waste generated during decommissioning can reduce the cost and optimize the space usage at disposal site. Therefore, efficient volume reduction is a very important factor during the decommissioning process. A conceptual method, based on the experiences of developing 200 and 800 ton compactors at Orion EnC, has been developed considering worker exposure with the followings a crusher (upgrade of compaction efficiency), an automatic dose measuring system with a NaI(Tl) detector, a shield box, an inner drum to prepare for easy handling of drums and packaging, a 30 ton compactor, and an automatic robot system. This system achieves a volume reduction ratio of up to 85.7%; hence, the system can reduce the disposal cost and waste volume. It can be applied to other types of wastes that are not easily managed due to high dose rates and remote control operation necessity.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4181-4190
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• 2023

Nuclear power plants are subjected to various processes during decommissioning, including cutting, decontamination, disposal, and treatment. The cutting of massive bio-shields is a significant step in the decommissioning process. Cutting is performed near the target structure, and during this process, workers are exposed to potential radioactive elements. However, studies considering worker exposure management during such cutting operations are limited. Furthermore, dismantling a nuclear power plant under certain circumstances may result in the unnecessary radiation exposure of workers and an increase in secondary waste generation. In this study, a cutting scenario was formulated considering the bio-shield as a representative structure. The specifications of a standard South Korean radioactive waste disposal drum were used as the basic conditions. Additionally, we explored the hot-to-cold and cold-to-hot methods, with and without the application of polishing during decontamination. For evaluating various scenarios, different cutting time points up to 30 years after permanent shutdown were considered, and cutting speeds of 1-10nullm²/h were applied to account for the variability and uncertainty attributable to the design output and specifications. The obtained results provide fundamental guidelines for establishing cutting methods suitable for large structures.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.15-28
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• 2019

In accordance with the decommissioning plan for the Kori Unit 1 NPP, the reactor coolant system will be chemically decontaminated as soon as possible after permanent shutdown. This study developed the chemical decontamination process though the development project of decontamination technology of reactor coolant system and dismantled equipment for NPP decommissioning, which has been carried out since 2014. In this study, Oxidation/reduction process was conducted using system decontamination process development equipment of lab scale and was divided into unit and continuous processes. The optimal process time was derived from the unit process, and decontamination agent and the number of process were derived through the continuous processes. Through the unit process, the oxidation process took 5 hours and the reduction process took 4 hours. As optimum decontamination agent, the oxidizing agent was $200mg{\cdot}L^{-1}$ Permanganic acid + $200mg{\cdot}L^{-1}$ Nitric acid and the reducing agent was $2000mg{\cdot}L^{-1}$ Oxalic acid. In the case of the number of processes, all oxide films were removed during the two-cycle chemical decontamination process of STS304 and SA508. In the case of Alloy600, all oxide films were removed when chemical decontamination was performed for three cycles or more.

• Proceedings of the Korea Society for Simulation Conference
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• 2000.11a
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• pp.61-66
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• 2000

The KEPRI has been upgrading the KNPEC(Korea Nuclear Power Education Center) #2 simulator, a replica of Yonggwang Unit 1 & 2, due to the outdated systems. The scenarios, such as the continuous load change, are selected to verify and validate the simulator, and the data required to V&V are generated with the best-estimated codes, RETRAN and MARS. The reactor coolant system and steam generator system are cut up into volumes and junctions for the accurate model of the scenarios, and other components and control systems are modeled. For the model the operation and design data of the plants is used and in some cases the data of Kori Unit 3 & 4 is used to fill up the lack of required data. The results of some selected analyses with the models are compared with the operating data of the plants to verify the models, and the analyses of the scenarios are carried out to generated the data for the V&V of the simulator

• Journal of Energy Engineering
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• v.12 no.2
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• pp.101-108
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• 2003

The purpose of this study is to develop an entrainment model of the Pressurized Water Reactor (PWR) U-tube Steam Generator (SG) for Main Steam Line Break (MSLB) analyses. Generally, the temperature of the inside containment vessel at MSLB is decreased by introducing the liquid entrainment effect. This effect makes a profit on the aspect of integrity evaluation for Equipment Environmental Qualification (EEQ) in the containment. However, the target plant, Kori unit 1 does not have the entrainment data. Therefore, this study has been performed. RETRAN-3D and LOFTRAN computer programs are used for the model development. There are several parameters that are used for the initial benchmark, such as Combustion Engineerings (CE) experimental data and the RETRAN-3D model which describes the test leg. A sensitivity study is then performed with this model in which the model parameters are varied until the calculated results provide reasonable agreement with the measured results for the entire test set. Finally, a multiplication factor has been obtained from the 95/95 values of the calculated (best-estimate) quality data relative to the measured quality data. With this new methodology, an additional temperature margin of about 40$^{\circ}C$ can be obtained. So, the new methodology is found to have an explicit advantage to EQ analyses.