• Nuclear Engineering and Technology
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• 제52권4호
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• pp.797-801
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• 2020

Most of irradiated graphite that should be disposed comes from moderators and reflectors of nuclear power plants. The quantity of irradiated graphite could be higher in the future if high-temperature reactors (HTRs) will be deployed. In this case noteworthy quantities of fuel pebbles containing semi-graphitic carbonaceous material should be added to the already existing 250,000 tons of irradiated graphite. Industry graphite is largely used in industrial applications for its high thermal and electrical conductivity and thermal and chemical resistance, making it a valuable material. Irradiated graphite constitutes a waste management challenge owing to the presence of long-lived radionuclides, such as ¹⁴C and ³⁶Cl. In the ENEA Nuclear Material Characterization Laboratory it has been successfully designed a procedure based on the exfoliation process organic solvent assisted, with the purpose of investigate the possibility of achieving graphite significantly less toxic that could be recycled for other purpose [1]. The objective of this paper is to evaluate the possibility of the scalability from laboratory to industrial dimensions of the exfoliation process and provide the prototype of a chemical plant for the treatment of irradiated graphite.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1037-1040
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• 2019

For the past 50 years, graphite has been widely used as a moderator, reflector and fuel matrix in different kinds of gas-cooled reactors. Resulting in approximately 250,000 metric tons of irradiated graphite waste. One of the most significant long-lived radioisotope from graphite reactors is carbon-14 ($^{14}C$) with a half-life of 5730 years, this makes it a huge concern for deep geologic disposal of nuclear graphite (NG). Considering the lifecycle of NG a number of waste management options have been developed, mainly focused on the achievement the radiological requirements for disposal. The existing approaches for recycling depend on the cost to be economically viable. In this new study, an affordable process to remove $^{14}C$ has been proposed using samples taken from the Nuclear Power Plant in Latina (Italy) which have been used to investigate the capability of organic and inorganic solvents in removing $^{14}C$ from exfoliated nuclear graphite, with the aim to design a practicable approach to obtain graphite for recycling or/and safety disposed as L& LLW.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.1209-1210
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• 2005

• 대한안전경영과학회:학술대회논문집
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• 대한안전경영과학회 2011년도 추계학술대회
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• pp.85-91
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• 2011

Safety in the nuclear facility has been a growing interest due to recent recurrences of the fatal accidents such as Fukushima accident and Chernobyl accident. It is not easy to determine the extent to what technical requirements of nuclear facility such as nuclear power plant are be likely applicable to the laboratory for use of nuclear material. All of workers in nuclear shall be recognized for the generic features of safety according to the related laws. This study surveys a status of safety control laws to enhance safety in laboratory for use of nuclear material.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1010-1018
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• 2017

The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2839-2846
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• 2021

In this study, experiments of downward-facing flow boiling were conducted to investigate material effects on CHF. Experiments were conducted using aluminum, copper, and carbon steel. It was found that different materials had different CHFs. Aluminum has the biggest CHF while copper has the lowest CHF for each mass flux. After experiment, surface wettability increased and surface became rougher, which was probably due to the oxidation process during nucleate boiling. The CHF difference is likely to be related to the surface wettability, roughness and thermal effusivity, which influences the bubble behavior and in turn affects CHF. Further studies are needed to determine which factor is dominant.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2874-2879
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• 2020

Due to excess reactivity, fresh nuclear fuel often contains burnable poisons. This research looks at six different burnable poisons and their impacts on reactivity, material attractiveness, and waste management. An MCNP simulation of a PWR fuel pin was performed with a fuel burnup of 60 GWd/MTHM to determine when each burnable poison fuel type would decrease below a k_∞ of 1. For determining the plutonium material attractiveness in each burnable poison fuel type, the plutonium isotopic content of the used fuel was evaluated using Bathke's Figure of Merit formula. For the waste management analysis, the thermal output of each burnable poison fuel type was determined through ORIGEN decay simulations at 100 and 300 years after being discharged from the core. The performance of all six burnable poisons varied over the three criteria considered and no single burnable poison performed best in all three considerations.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.803-810
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• 2013

The U.S. Department of Energy's Fuel Cycle Technologies R&D program under the Office of Nuclear Energy is working to advance technologies to enhance both the existing and future fuel cycles. One thrust area is in developing enabling technologies for next generation nuclear materials management under the Materials Protection, Accounting and Control Technologies (MPACT) Campaign where advanced instrumentation, analysis and assessment methods, and security approaches are being developed under a framework of Safeguards and Security by Design. An overview of the MPACT campaign's activities and recent accomplishments is presented along with future plans.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.23-24
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• 2017

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.1022-1028
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• 2020

One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different ²³⁵U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.