• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.365-369
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• 2022

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate-level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Journal of the Korean Society of Radiology
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• v.12 no.2
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• pp.139-148
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• 2018

Assessment of NPP(Nuclear Power Plant) decommissioning cost is very important for safe decommissioning of nuclear power plants. In the United States, which has the most NPP decommissioning experience, the cost evaluation study has been conducted since the 1970s in order to decommissioning nuclear facilities. The US NRC has conducted studies on decommissioning technology, safety and cost for a variety of reactor type and nuclear installations. In the total decommissioning costs, the end of operation licenses accounted for the largest portion, followed by spent fuel management and site restoration. In case of immediate decommissioning, spent fuel management cost increased compared to delayed decommissioning, and delayed deocmmissioning increased the cost of terminating the operation license. However, in general, delayed decommissioning does not show any significant benefit as compared with immediate decommissioning. It is necessary to consider the evaluation according to the site conditions when evaluating the cost of decommissioning domestic nuclear power plants. Also, in Korea, IAEA recommendations were applied to reorganize the radioactive waste classification system. Therefore, it is necessary to develop a method to appropriately use the decommissioning data of the preceding US Nuclear Power Plant in the new classification system when estimating the amount of radioactive waste generated during decommissioning. In particular, the establishment of the evaluation methodology for the waste to be disposed of will be an important factor in securing the accuracy of the decommissioning cost. In addition, it is necessary to construct information data that can be applied to facility characteristics and work characteristics in order to evaluate the cost of demolition of domestic nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.219-220
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• 2005

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1491-1510
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• 2021

This work presents an illustration of Nagra's AMAC (Advanced Methodology for Activation Characterization) approach to the South Korean pressurized water reactor Kori-1 decommissioning. The results achieved are supported by the existing experience from the detailed AMAC applications to Swiss NPPs and are used not only for a demonstration of the applicability of AMAC to South Korean NPPs, but also for a first approximation of the activated waste volumes to be expected from Kori-1. A packaging concept based on the above activation characterization is also presented, using the AMAC algorithmic optimization software ALGOPACK leading to the minimum number of waste containers needed given the selected packaging constraints. Nagra's AMAC enables effective planning before and during NPP decommissioning, including recommendations for cutting profiles for diverse reactor components and building structures. Finally, it is expected to lead to significant cost savings by reducing the number of expensive waste containers, by optimizing a potential melting strategy for metallic waste as well as by significantly limiting the number of radiological measurements. All information about Kori-1 used for the purpose of this study was collected from publicly available sources.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.69-75
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• 2004

DORT and ORIGEN2 code were used for calculation of neutron flux and inventory in reactor pressure vessel(RPV) of Kori unit-1, To calculate neutron flux using DORT code, the reactor was divided into 94 mesh from the center of core to RPV and from 0 to 45 degree along the azimuth. The cross-sections of main nuclides were recalculated using neutron flux in the RPV region. The results showed that 95% of the total activity in RPV came from the nuclides of $^{55}$ Fe, $^{60}$ Co, $^{59}$ Ni and $^{63}$ Ni. And the total activity with cooling of more than 50 years after decommissioning was no more than 0.2% of at the time of shutdown. Considering the weight of RPV is 210 tons, the initial total activity of RPV reached 5.25${\times}$10$^{6}$ GBq. To verify results of ORIGEN2 calculation, comparison between calculated and measured value at RPV of Kori unit-1 was peformed. The comparison results showed a good agreement.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.443-448
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• 2006

The OPAL (Open Pool Australian Light-water) reactor is currently being constructed to replace HIFAR (HI-Flux Australian Reactor, commissioned in 1958) in mid-2006. HIFAR will be shutdown for decommissioning after several months of simultaneous operation with OPAL for smooth transition of operating systems and business. OPAL is a 20 MW multipurpose research reactor for radioisotope production, irradiation services and neutron beam research. The OPAL reactor uses low enriched uranium fuel in a compact core, cooled by light water and moderated by heavy water, yielding maximum thermal flux not less than $4{\times}10^{14}ncm^{-2}s^{-1}$. The reactor containment building is constructed of reinforced concrete and has been designed to protect the reactor from all external events such as seismic occurrences and impact from a hypothetical light aircraft crash. This paper describes the main elements of the reactor design and its applications.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.1
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• pp.63-71
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• 2018

This paper discusses decommissioning procedure requirements management using requirement engineering to systematically manage the technical requirements and criteria that are required in decontamination and decommissioning activities, and the regulatory requirements that should be complied with in a decommissioning strategy for research reactors and nuclear power plants. A schema was designed to establish the traceability and change management related to the linkage between the regulatory requirements and technical criteria after classifying the procedures into four groups during the full life-cycle of the decommissioning. The results confirmed that the designed schema was successfully traced in accordance with the regulatory requirements and technical criteria required by various fields in terms of decontamination and decommissioning activities. In addition, the changes before and after the revision of the Nuclear Safety Act were also determined. The dismantling procedure requirement management system secured through this study is expected to be a useful tool in the integrated management of radioactive waste, as well as in the dismantling of research reactor and nuclear facilities.

• Journal of Radiation Protection and Research
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• v.45 no.4
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• pp.178-186
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• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.105-114
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• 2004

This study investigated the thermal behavior of the nuclear graphite waste generated from the decommissioning of the Korean nuclear research reactor, The first part study investigated the decomposition rate of the nuclear graphite waste up to $1000^{\circ}C$ under various oxygen partial pressures using a thermo-gravimetric analyzer (TGA). Tested graphite waste sample not easily destroyed in the oxygen-deficient condition. However, the gas-solid oxidation reaction was found to be very effective in the presence of oxygen. No significant amount of the product of incomplete combustion was formed even in the limited oxygen concentration of 4% $O_2$. The influence of temperature and oxygen partial pressure was evaluated by the theoretical model analysis of the thermo-gravimetric data. The activation energy and the reaction order of graphite oxidation were evaluated as 128 kJ/mole and 1.1, respectively. The second part of this study investigated the behavior of radioactive elements under graphite oxidation atmosphere using thermodynamic equilibrium model. $^{22}Na$, $^{134}Cs$ and $^{137}Cs$ were found be the semi-volatile elements. Since volatile uranium species can be formulated at high temperatures above $1050^{\circ}C$, the temperature of incinerator furnace should be minimized. Other corrosion/activation products, fission products and uranium were found to be the non-volatile species.