• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.219-230
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• 2017

This study was conducted to predict and evaluate the uncertainty of safety after closure of the second phase surface disposal facility of the Gyeongju intermediate and low level repository in Korea. In this study, four scenarios are developed considering both intact and degraded states of multi-layered covers and disposal containers; also, the fluid flow by a rainfall into the disposal facility is simulated. The rainfall conditions are implemented based on the monthly average data of the past 30 years (1985~2014); the simulation period is 300 years, the management period regulated by institutional provisions. As a result of the evaluation of the basic scenario, in which the integrity of both of the containers and the covers is maintained, it was confirmed that penetration of rainfall does not completely saturate the inside of the disposal facility. It is revealed that the multiple cover layers and concrete containers effectively play the role of barrier against the permeation of rainfall.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.43-63
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• 2022

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1^st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl₂ and CrCl₃, other metal chloride species such as FeCl₂, FeCl₃, MoCl₂, MnCl₂ and NiCl₂ will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of < 1 ppm per day. During the incubation period, metal concentration show 1^st order kinetics and increasing linearly with time^1/2. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.381-390
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• 2017

In Korea, commercial nuclear power plants and research reactors have on-site storage systems for the spent nuclear fuel, but it is difficult to expand the facilities used for the storage systems. If decommissioning of nuclear power plants starts, an amount of high level radioactive waste will be generated. In this study, a radiological impact assessment of the railroad transport of high level radioactive waste was carried out considering radiation workers and the public, using the developed transport container as the transport package. The dose rates for workers and the public during the transport period were estimated, considering anticipated transport scenarios, and the results compared with the regulatory limit. A sensitivity analysis was also carried out by considering the different release ratios of the radioactive materials in the high level radioactive waste, and different distances between the transport container and workers during loading and unloading phases and while attaching another freight car. For all the anticipated transport scenarios, the radiological impacts for workers and the public met the regulatory limits.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.93-98
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• 2011

Saline corrosion is one of the major degradation mechanisms for stainless steel type 304 (SS304) dry storage cask during the spent fuel interim storage period. Slow strain rate test (SSRT) and neutral salt spray test (NSS) were performed at $85^{\circ}C$ and $200^{\circ}C$ with 0.5 wt% sodium chloride mist sprayed on the cold-rolled SS304 specimens of different degrees of reduction in this study. The weight changes of the NSS specimens tested at $85^{\circ}C$ for 2000 hours differed greatly from those at $200^{\circ}C$. The weight loss of NSS specimens was not significant at $85^{\circ}C$ but the weight gain decreased gradually with increasing the cold-rolled reduction. The yield strength (YS) and ultimate tensile stress (UTS) values obtained from the SSRT tests for lightly cold-rolled specimens in the salt spray environment at $85^{\circ}C$ and $200^{\circ}C$ are slightly lower than in air. But for those with 20% reductions, the specimen strengths were no longer changed by the saline corrosion. The preliminary results demonstrated that the quality and performance of cold-rolled SS304 is acceptable for fabrication of dry storage casks. However, more work on the corrosion behavior of cold-rolled stainless steel in the saline atmosphere is needed to better understand its long-term performance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.247-260
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• 2020

A multi-regression model was developed to estimate the decommissioning cost for Kori unit 1 using foreign nuclear power plant (NPP) decommissioning cost data. First, the decommissioning cost data were collected for 13 boiling water reactors and 16 pressurized water reactors and converted into the values as of November 2019. Then, for the regression model, the decommissioning cost was chosen as the dependent variable, and two variables were selected as independent variables: a contamination factor that was designed to reflect the operational characteristics of the decommissioned NPP and the decommissioning period. A statistical package in the R language was used to derive the regression model. Finally, the regression model was applied to estimate the decommissioning cost for Kori unit 1. The estimated decommissioning cost for Kori unit 1 was 663.40~928.32 million US dollars (782,812~1,095,418 million Korean won).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.105-112
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• 2004

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.189-200
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• 2016

The Republic of Tajikistan has ten former uranium mining sites. The total volume of all tailings is approximately 55 million tonnes, and the covered area is more than 200 hectares. The safe management of legacy uranium mining and tailing sites has become an issue of concern. Depending on the performance requirements and site-specific conditions (location in an arid, semiarid or humid region), a cover system for uranium tailings sites could be constructed using several material layers using both natural and man-made materials. The purpose of this study is to find a feasible cost-effective cover system design for the Degmay uranium tailings site which could provide a long period (100 years) of protection. The HELP computer code was used in the evaluation of potential Degmay cover system designs. As a result of this study, a cover system with 70 cm thick percolation layer, 30 cm thick drainage layer, geomembrane liner and 60 cm thick barrier soil layer is recommended because it minimizes cover thickness and would be the most cost-effective design.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.329-346
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• 2008

Post-closure safety assessment for the Wolsong Low- and Intermediate-level radioactive waste Disposal Center is described. Based on assessment context, closure concept and ground water flow characteristics of the disposal site, brief descriptions are included on the assessment scenarios, models, input parameters and tools. Radionuclide transport modeling in the near-field and far-field, gas generation and transport modeling, human intrusion and biosphere transport are also described briefly. Assessment results for each scenarios are shown to meet the performance criteria of regulatory body. Further and continuous efforts to improve the safety of disposal facility will be made during the construction and operational period.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.229-238
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• 2010

In order to acquire a realistic forecast for the lifetime and post-closure period of the LILW (Low- and Intermediate-Level Radioactive Waste) repository and to establish the overall management plan associated gas issues. it is essential to carry out the long-term experimental research in a similar condition to actual disposal environment. Regarding this, as a part of the following-up actions on a construction and operation license for the first stage of the LILW repository at Gyeongju city, a large-scale in-situ experiment is being planned. For securing basic data on the experiment, the experimental researches related to gas generation previously performed in foreign countries are reviewed in detail. Consequently, it is judged that data on the gas generation experiment in Finland could be practically applied as the benchmark for our large-scale in-situ experiment because the same disposal concept as the Korean repository is adopted and the experiment is performed in a scale large enough to allow the use of regular waste packages.