• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.386-395
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• 2004

The Korean fuel cycle scenario has been modeled by using the dynamic analysis method. For once-through fuel cycle model, the nuclear power plant construction plan was considered, and the nuclear demand growth rate from the year 2016 was assumed to be 1%. After setup the once-thorough fuel cycle model, the DUPIC and fast reactor scenarios were modeled to investigate the environmental effect of each fuel cycle. Through the calculation of the amount of spent fuel, and the amounts of plutonium and minor actinides were estimated and compared to those of the once-through fuel cycle. The results of the once-through fuel cycle shows that the demand grows to 64 GWe and the total amount of the spent fuel would be 100 kt in the year 2100, while the total spent fuel can be reduced by 50% when the DUPIC scenario is implemented

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.359-367
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• 2018

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.191-195
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• 2001

An accurate and simple analytical technique for low levels of fallout Pu in the environment was developed using an anion exchange resin. To develop the reliable determination of Pu isotopes in soil samples, decomposition of soil matrix, plutonium oxidation state adjustment on the anion exchange column and source preparation of Pu isotopes have been carried out. The optimum method of Pu isotopes with an anion exchange has been validated by application to IAEA-Reference soils.

• Journal of the Korean Chemical Society
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• v.44 no.5
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• pp.480-484
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• 2000

• Nuclear Engineering and Technology
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• v.13 no.3
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• pp.139-144
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• 1981

Plutionum, neptunium, and uranium isotopes which were produced in the interaction of $^{136}$ Xe and $^{244}$ Pu are separated and determined their cross sections. The cross sections are: $\sigma$($^{245}$ Pu)=66 mb, $\sigma$($^{243}$ Pu)=57 mb, $\sigma$($^{246}$ Pu)=6.0 mb, $\sigma$($^{239}$ Np)=15 mb, $\sigma$($^{240}$ U)=12 mb, $\sigma$($^{245}$ U)=6.4 mb respectively.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.194-213
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• 1980

The paper describes and quantifies the wastes arising in the nuclear fuel cycle for Light Water Reactors, Heavy Water Reactors and Fast Breeder Reactors. The management and disposal technologies are indicated, together with their environmental impacts. Both once-through and uranium-plutonium recycle systems are evaluated, and comparisons are made on the basis of tingle reference technologies for waste management, and for one gigawatt/year of electricity generation. Environmental impacts are assessed, particularly that of health and safety, and a reference costing system is applied purely as a basis for comparing the fuel cycles. From this study it call be concluded generally that the relative differences of the impacts of waste management and disposal between the selected fuel cycles are not decisive factors in choosing a fuel cycle. Employing the technologies assumed, the radioactive wastes from any of the fuel cycles studied can be managed and disposed of with a high degree of safety and without undue risk to man or the environment. The cost of waste management and disposal is only a few percent of the value of the electricity generated and does not vary greatly between fuel cycles.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.46-51
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• 1996

In this study we calculate the infite multiplication factors ($k_{\infty}$) and the Doppler temperature coefficients (DTC) of two mixed-oxide (MOX) fuel rods with different plutonium contents by using PHOENIX-P, HELIOS and CASMO-3 codes. The results were compared against the reference values obtained by MCNP-3A continuous-energy Monte Carlo code. The purpose of this study is to benchmark the accuracy of these lattice codes. The PHOENIX-P's Doppler coefficients calculated were in good agreement with the MCNP results within the Monte-Carlo uncertainty band which is in the order of $\pm$ 10% for the Doppler coefficients..

• Nuclear Engineering and Technology
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• v.8 no.4
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• pp.219-229
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• 1976

A statistical approach is employed to investigate the relative advantages of several alternative fuel cycles suitable for a hypothetical 1125 MWe plant in Korea. All the fuel cost parameters are treated as statistical variables, each being associated with an appropriate probability distribution function. Through a random sampling procedure, the probability histograms on both capital requirements and break-even costs of various fuel cycle components are obtained. The histograms are then utilized to quantify the cost-benefit of the fuel cycle with reprocessing or the plutonium recycle over the throwaway cycle.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2339-2358
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• 2022

The review discusses various alternative approaches for spent nuclear fuel (SNF) reprocessing in aqueous carbonate media. The main stages, schemes, and methods of the most well-known and well-described processes for reprocessing SNF and some high-level radioactive waste using carbonate systems developed by research groups in Japan, the United States of America, the Republic of Korea, and the Russian Federation described and compared. The main advantages of such methods are outlined compared to the SNF reprocessing in nitric acid media. The levels of development and proximity of the designed processes to the industrial implementation are shown. The main principle achievements, prospects, and routes for the refinement of such methods for the technology of SNF reprocessing and handling of high-level radioactive waste formulated.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.391-407
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• 2023

This article provides a survey of wet (aqueous) methods for recovery, separation, and purification of uranium from fission products in carbonate solutions during the reprocessing of spent nuclear fuel and methods for removal of radionuclides from alkaline radioactive waste. The main methods such as selective direct precipitation, ion exchange, and solvent extraction are considered. These methods were compared and evaluated for reprocessing of spent nuclear fuel in carbonate media according to novel alternative non-acidic methods and for treatment processes of alkaline radioactive waste.