• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.12-18
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• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population. Many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solve the materials aging problem is integral to its success. A foundation for effective aging management of nuclear power plants is that aging is properly taken into account at each stage of a plant's lifetime, i.e. in design, manufacture, construction and operation including long term operation and decommissioning. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a regulation on the integrated materials aging management for nuclear facilities is proposed. The proposed regulation identifies key elements of effective aging management for nuclear power plants and provides the requirements on aging management for nuclear facilities throughout all stages of the lifetime of the plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.79-86
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• 2009

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.21 no.1
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• pp.22-27
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• 2022

In accordance with the growing trend of decommissioning nuclear facilities, research on the cutting process is actively proceeding worldwide. In general, a thermal cutting process, such as plasma cutting is applied to decommissioning a nuclear reactor pressure vessel (RPV). Plasma cutting has the advantage of removing the radioactive materials and being able to cut thick materials. However, when operating under water, the molten metal remains in the cut plane and re-solidifies. Hence, cutting is not entirely accomplished. For these environmental reasons, it is difficult to cut thick metal. The contact arc metal cutting (CAMC) process can be used to cut thick metal under water. CAMC is a process that cuts metal using a plate-shaped electrode based on a high-current arc plasma heat source. During the cutting process, high-pressure water is sprayed from the electrode to remove the molten metal, known as rinsing. As the CAMC is conducted without using a shielding gas, such as Argon, the electrode is consumed during the process. In this study, CAMC is introduced as a method for dismantling nuclear vessels and the relationship between the metal removal and electrode consumption is investigated according to the cutting conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2806-2820
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• 2024

Accurate reconstruction of radiation field and path planning are very important for the safety of operators in the process of dismantling nuclear facilities. Based on radial basis function (RBF) interpolation algorithm, this paper discussed the application of inverse multiquadric radial basis Function (IMRBF) interpolation method to the reconstruction of gamma radiation field, and proved the feasibility of reconstructing a radiation field with multiple γ sources. The average relative errors of IMRBF interpolation results were 4.28% and 8.76%, respectively, for the experimental scenarios with single and double gamma sources. After comparing the consistency between the simulated scene and the experimental scene, IMRBF method and Cubic Spline method were respectively used to reconstruct the gamma radiation field by Geant4 simulation data. The results showed that the interpolation accuracy of IMRBF method was superior to that of Cubic Spline method. Further, more RBF interpolation algorithms were used to reconstruct the multi-γ source radiation field, and then the Probabilistic Roadmap (PRM) algorithm was used to optimize the human walking path in the radiation field reconstructed by different interpolation methods. The optimal paths in radiation fields generated by multiple interpolation methods were compared. The results herein contribute to a comprehensive understanding of RBF interpolation methods in reconstructing γ radiation fields and their application in optimizing paths in radiation environments. The insights may provide valuable information for decision-making in radiation protection during the decommissioning of nuclear facilities.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.633-646
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• 2020

Imaging and scattering techniques using thermal neutrons allow to analyze complex specimens in scientific and industrial researches. Owing to this advantage, there have been a considerable demand for neutron facilities in the industrial sector. Among neutron sources, an accelerator driven compact neutron source is the only one that can satisfy the various requirements-construction budget, facility size, and required neutron flux-of industrial applications. In this paper, a target, moderator, and reflector (TMR) system for low-energy proton-accelerator driven compact thermal neutron source was designed via Monte Carlo simulations. For 10-30 MeV proton beams, the optimal conditions of the beryllium target were determined by considering the neutron yield and the blistering of the target. For a non-borated polyethylene moderator, the neutronic properties were verified based on its thickness. For a reflector, three candidates-light water, beryllium, and graphite-were considered as reflector materials, and the optimal conditions were identified. The results verified that the neutronic intensity varied in the order beryllium > light water > graphite, the compacter size in the order light water < beryllium < graphite and the shorter emission time in the order graphite < light water < beryllium. The performance of the designed TMR system was compared with that of existing facilities and were laid between performance of existing facilities.

• International conference on construction engineering and project management
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• 2020.12a
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• pp.353-358
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• 2020

Recent efforts to develop a common standard for nuclear power plants (NPPs) with the aim of creating (1) a digital environment for a better understanding of NPPs life-cycle management aspect and (2) engineering data interoperability by using existing standards among different unspecified project participants (e.g., owners/operators, engineers, contractors, equipment suppliers) during plants' life cycle process (EPC, O&M, and decommissioning). In order to meet this goal, there is a need for formulating a standardized high-level physical breakdown structure (PBS) for NPPs project management office (PMO). However, high-level PBS must be comprehensive enough and able to represent the different types of plants and the new trends of technologies in the industry. This has triggered the need for addressing the issues of the recent operational NPPs and future technologies' ramification for evaluating the changes in the NPPs physical components in terms of structure, system, and component (SSC) configuration. In this context, this ongoing study examines the recent conventional NPPs and technological trends in the development of future NPPs facilities. New reactor models regarding the overlap of variant issues of nuclear technology were explored. Finally, issues on PBS for project management are explored by the examination of the configuration of NPPs primary system. The primary systems' configuration of different reactor models is assessed in order to clarify the need for analyzing the new trends in nuclear technology and to formulate a common high-level PBS. Findings and implications are discussed for further studies.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1368-1375
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• 2016

A radiological safety assessment study was performed for the transportation of low level radioactive wastes which are temporarily stored in Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea. We considered two kinds of wastes: (1) operation wastes generated from the routine operation of facilities; and (2) decommissioning wastes generated from the decommissioning of a research reactor in KAERI. The important part of the radiological safety assessment is related to the exposure dose assessment for the incidentfree (normal) transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public. The effective doses were estimated based on the detailed information on the transportation plan and on the radiological characteristics of waste packages. We also estimated radiological risks and the effective doses for the general public resulting from accidents such as an impact and a fire caused by the impact during the transportation. According to the results, the effective doses for transport personnel, radiation workers, and the general public are far below the regulatory limits. Therefore, we can secure safety from the viewpoint of radiological safety for all situations during the transportation of radioactive wastes which have been stored temporarily in KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.93-105
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• 2018

In Korea, a huge amount of radioactive concrete waste will be generated through decommissioning of nuclear facilities in the near future; therefore, optimum technology for the treatment of concrete waste should be reviewed thoroughly and the future direction of technology development should be discussed. In this paper, many domestic and foreign examples of generation of radioactive concrete waste were pieced together and the characteristics of radioactive concrete waste were examined. Moreover, we reviewed trends in technology development by analyzing the examples of various studies and practical applications of treatment technologies, such as mechanical decontamination, chemical decontamination, volume reduction, recycling and solidification, and also tried to understand the limitations of existing technologies and determine a direction for technical improvement.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1079-1086
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• 2009

A great volume of radioactive concrete is generated during the operation and the decommissioning of nuclear facilities. The washing-electrokinetic technology in this study, which combined an electrokinetic method and a washing method, was developed to decontaminate the concrete generated in nuclear facilities. The results of only an electrokinetic decontamination for the concrete showed that cobalt was removed to below 1% from the concrete due to its high pH. Therefore, the washing-electrokinetic technology was applied to lower the pH of the concrete. Namely, when the concrete was washed with 3 M of hydrochloric acid for 4 hours (0.17 day), the $CaCO_3$ in the concrete was decomposed into $CO_2$ and the pH of the concrete was reduced to 3.7, and the cobalt and cesium in the concrete were removed by up to 85.0% and 76.3% respectively. Next, when the washed concrete was decontaminated by the electrokinetic method with 0.01M of acetic acid in the 1L electrokinetic equipment for 14.83 days, the cobalt and the cesium in the concrete were both removed by up to 99.7% and 99.6% respectively. The removal efficiencies of the cobalt and cesium by 0.01M of acetic acid were increased more than those by 0.05M of acetic acid due to the increase of the concrete zeta potential. The total effluent volume generated from the washing-electrokinetic decontamination was 11.55L (7.2ml/g).