• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.121-126
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• 2019

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, $^{36}Cl$ and $^{41}Ca$ are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of $^{36}Cl$ and $^{41}Ca$ in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of $^{36}Cl$ and $^{41}Ca$ using the activation calculation code ORIGEN2.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.17-26
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• 1986

A test is made for a method to determine reliable bias in the criticality safety analysis of spent fuel storage pool with turnup credit between the reference and rack criticality calculation methods. The spent fuel pool of Kori Unit 1 is conceptually redesigned to the most compact rack with turnup credit, and its multiplication factors are calculated depending on fuel enrichment and burnup, by the Monte Carlo code KENO-IV as a reference and by a two-dimensional collision probability code FATAC as a practical method. Then, the computed values with the help of the above two computer codes are compared to evaluate the bias and its trend in terms of multiplication factor on fuel enrichment and turnup. The result indicates that the bias can be determined with reliability basis but without any disadvantage in criticality safety margin compared with the conventional method.

• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.41-51
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• 1988

In order to develop a food-chain computer code suitable to the environmental conditions of Korea, the FOOD III code was partially modified. The excution results for Korean male-adult were compared to those from the Canadian version of FOOD III to deduce a more realistic approach in dose assessment. The amounts of Mn-54, Co-58, Co-60, I-131 and I-132 released from Kori unit 1 in1984 were used as the source terms for the sample calculation. The maximum atmospheric dispersion factor(X/Q) value on the site boundary was applied. Through the code modification, organ doses decreased by about $20{\sim}70%$ and the effective committed dose equivalent by about 40% to be $7.935{\times}10^{-6}Sv/y$ which is 0.16% of the ICRP limit, $5{\times}10^{-3}Sv/y$.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.234-243
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• 2021

In the research or project planning for the decommissioning of a nuclear power plant, one of several preparations will be the establishment of a list of potential radionuclides to be considered at the time of characterization or final status surveys. Reliable data for selection of potential radionuclides during the transition period to prepare for decommissioning will depend heavily on historical data at the site or, where possible, sampling analysis. However, during the transition period, direct sampling can be challenging, depending on the circumstances of the site or national regulation. A methodology of selecting potential radionuclides for nuclear facility sites which largely consists of three major processes: production of initial list of radionuclides, selection of the insignificant radionuclide that will be eliminated, and consideration of site characterization or sampling. For developing a preliminary list of potential radionuclides for Kori Unit 1 decommissioning, the list of initial radionuclides was made referring to the technical documents applied at decommissioned NPPs in the U.S and additional reference materials applied until the operation of NPPs in Korea. For the screening of insignificant radionuclides, we applied criterion of less than 0.1% of the amount of radioactivity inventory and confirmed the dose fraction using the RESRAD code. The final suit of radionuclides was established, which should be supplemented by reflecting site characterization and sampling process in the future. Thus, the methodology and results for the selection of potential radionuclides suggested in this paper can give an insight as a future reference to deriving DCGLs in relation to site remediation of decommissioning nuclear plants.

• Environmental and Resource Economics Review
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• v.27 no.2
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• pp.261-286
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• 2018

This paper evaluated the economic feasibility of the life extension of Kori unit 1 and Wolsong unit 1 according to the types of the nuclear power plants (NPPs) and the life extension period comparing to the levelized costs of energy (LCOE) of the new NPPs, coal-fired plants (CFPs), and combined cycle gas turbine (CCGTs) which proposed in the $7^{th}$ Basic Plan for Electricity Supply and Demand. The economic feasibility of the life extension of NPPs using LCOE method is affected by the types of NPPs, lifetime extension periods, discount rate, and capacity factor. According to the analysis results, the pressurized light water reactor (PWR) is more economical than the pressurized heavy water reactor (PHWR). Comparing the economical efficiency between the life extension of NPPs and other alternatives, the operation of the PWR for 20 years is more economical than the one of new NPPs and CFPs. However, 20 years of life extension of PHWR is more economical than the CCGTs, but less economical than new NPPs and CFPs. In summary, the 20 years of life extension of the NPPs seems to be more, especially for the PWR, which is more cost effective than other generation alternatives. Therefore, the government policy of the life extension of NPPs need to be a selective approach that simultaneously considers both safety and economics rather than closing all NPPs.

• Journal of Radiation Protection and Research
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• v.9 no.1
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• pp.26-32
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• 1984

The exposure doses by the radioactive gaseous effluents from nuclear power plants are investigated in the two cases of normal operation and hypothetical accident. Gaussian equation is adapted in the normal operation as the diffusion model of effluents for long period, which uses annual average meteorological data. But the real time models have been used in the case of accidents which analyze the changes of wind direction and speed. In this study the annual exposure doses by the normal operation of Kori unit 1 during $1977{\sim}1982$ were calculated on the basis of the atmospheric diffusion factor by the Gaussian straight line model. And the image processing technique was suggested as the effective method through the wind tunnel experiments to get the characteristic value of atmospheric diffusion coefficient required especially in the accidents of nuclear power plants.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.98-103
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• 2009

The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.355-367
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• 2017

The decommissioning of Kori unit 1, which was permanently shut down in June of 2017, will be the first instance of the dismantling of a commercial nuclear power plant in Korea. The disposal of waste during the dismantling process accounts for a large part of the total decommissioning cost. Therefore, structures consisting of activated and contaminated concrete must be economically and safely dismantled by establishing a proper dismantling strategy. This study focuses on optimized dismantling and disposal scenarios pertaining to a biological shield. Several dismantling cases, regulations and technologies related to waste treatment as these practices pertain to nuclear power plants are analyzed. To minimize the amount of waste from the biological shield dismantling process, an optimized dismantling scenario is presented and disposal alternatives for dismantled concrete waste are proposed.

• Journal of the Korean Recycled Construction Resources Institute
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• v.7 no.4
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• pp.316-322
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• 2019

A probabilistic approach considering uncertainties was employed to investigate the effects on the material characteristics and strength of nuclear bio-logical shield concrete walls, when exposed to long-term radiation during the entire service life. Time-dependent compressive and tensile strengths were estimated by conducting the neutron fluence analysis. For the contaminated concrete, individual compressive and tensile failure probabilities can be possibly evaluated by not only establishing limit-state function withthe predefined critical values but also performing Monte Carlo Simulation. Nuclear power plant types similar to the Kori Unit 1, which was shut off permanently in 2017 after the 40-year operation, were herein selected for an illustrative purpose. Consequently, the probability-based performance assessment and prediction of contaminated concrete walls were well demonstrated.