• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2809-2818
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• 2022

In this investigation work, liquid-liquid extraction (L.L.E) through three distinctive frameworks have been examined for assurance of slip velocity (S.V), and characteristic velocity (C.V) in a non-center rotating column (N.C.R.C) with a wide extend of factors. Three double frameworks with distinctive interfacial tension comprising of toluene-water (high interfacial tension), n-butyl acetate-water (medium interfacial tension), and n-butanol-water (low interfacial tension) were investigated for tests. Two common relationships for the expectation of S.V and C.V, including phase stream rates, rotor speed, column geometry additionally physical properties, are displayed. The recommended relationships were compared with test information gotten from the writing and the display examination. Findings of this study, the present proposed correlations are more accurate than those previously reported.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.73-73
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• 2009

There are many factors to evaluate nuclear fuel cycle such as safety, public acceptance, economics, etc.. Transparency, proliferation, environment issues, public acceptance and safety are essential to expansion of nuclear industry and proliferation resistance is one of key constraints in the deployment of advanced nuclear energy systems. Proliferation resistance is being considered as one of the most important factors in assessing advanced and innovative nuclear systems. IAEA defmes proliferation resistance as characteristics of nuclear energy system that impedes the diversion or undeclared production of nuclear material [1]. Barriers to proliferation is consist of intrinsic and extrinsic barriers(institutional measures). Intrinsic barriers are characterized in material barriers and technical barriers in general. Material barriers is intrinsic, or inherent, qualities of materials that reduce the inherent desirability or attractiveness of the material as an explosive. Isotopic, chemical, radiological, mass and bulk, detectability barriers are considered as material barriers attributes [2]. Proliferation resistance is examined for several nuclear fuel cycles based on previous study which is focused on the intrinsic barriers [3-4]. Pyroprocessing and DUPIC are considered as reprocessing technologies in Korea and the PWR direct disposal is considered. Comparative assessments of the proliferation attributes and merits of different fuel cycle systems will be performed and the optimal back-end fuel cycle and strategy will be proposed.

• Journal of the Korean Professional Engineers Association
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• v.33 no.3
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• pp.15-19
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• 2000

The importance of the nuclear energy as one of the national energy sources, safety consideration of the nuclear power plants and status of nuclear power generation and nuclear fuel cycle in Korea are discussed.

• Journal of Institute of Convergence Technology
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• v.3 no.2
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• pp.1-6
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• 2013

The ionic liquids are known to potential alternative solvents capable of replacing the commercial solvents in various processes including those in nuclear fuel cycle. As to the material, a number of studies have already reviewed the interesting results and addressed the spectroscopic as well as electrochemical behaviors of metal elements included in spent nuclear fuels. It has found that the important properties of metal ions in TBP dissolved ILs have led the development of alternative technologies to traditional solvent extraction processes. On the other hand, the electrochemical deposition of metal ions in ILs have been investigated for the application of the solvents to aqueous as well as to non-aqueous processes. In this work, a review on the application of ILs in nuclear fuel cycle is briefly presented to understand the notable researches on ILs focusing on aqueous processes.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.56-66
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• 2005

In response to the increasing emphasis being placed on the importance of international cooperation as part of global efforts to cope with growing non proliferation, and security, concerns in the nuclear field, the Director General of the International Atomic Energy Agency (IAEA), Mohamed ElBaradei, appointed an international group of experts to consider possible multilateral approaches to the nuclear fuel cycle. The mandate of the Expert Group was three fold: ${\bullet}$ To identify and provide an analysis of issues and options relevant to multilateral approaches to the front and back ends of the nuclear fuel cycle; ${\bullet}$ To provide an overview of the policy, legal, security, economic, institutional and technological incentives and disincentives for cooperation in multilateral arrangements for the front and back ends of the nuclear fuel cycle; and ${\bullet}$ To provide a brief review of the historical and current experiences and analyses relating to multilateral fuel cycle arrangements relevant to the work of the Expert Group. The overall purpose was to assess MNAs in the framework of a double objective: strengthening the international nuclear non proliferation regime and making the peaceful uses of nuclear energy more economical and attractive. The Group identifies options for MNAs - options in terms of policy, institutional and legal factors - for those parts of the nuclear fuel cycle of greatest sensitivity from the point of view of proliferation risk. It also reflects the Groups deliberations on the corresponding benefits and disadvantages (pros and cons) of the various options and approaches. Although the Expert Group was able to agree to forward the resulting report to the Director General, it is important to note that the report does not reflect agreement by all of the experts on any of the options, nor a consensus assessment of their respective value. It is intended only to present options for MNAs, and to reflect on the range of considerations which could impact on the desirability and feasibility of those options.

• Nuclear Engineering and Technology
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• v.7 no.4
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• pp.295-310
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• 1975

A system of model price data for the fuel cost estimation of the Go-ri plant is developed. With the application of MITCOST-II computer code the levelized unit fuel costs over the entire lifetime of the plant are evaluated. It is found that the overall levelized unit fuel cost is 7.332 mills/Kwhe and that the uranium ore and enrichment service represent more than 85% of the unit cost, assuming a simple once-through fuel cycle process with no reprocessing of the spent fuel. The effects of the cost fluctuations in these fuel cycle elements and the capacity factor changes are also evaluated. The results indicate that the fuel costs are most sensitive to the variation of uranium ore price. Efforts must, therefore, be employed for the arrangement of cheap and timely supply of uranium ore in order to achieve the economic generation of nuclear power.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.647-661
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• 2015

The supercritical $CO_2$ (S-$CO_2$) Brayton cycle has recently been gaining a lot of attention for application to next generation nuclear reactors. The advantages of the S-$CO_2$ cycle are high efficiency in the mild turbine inlet temperature region and a small physical footprint with a simple layout, compact turbomachinery, and heat exchangers. Several heat sources including nuclear, fossil fuel, waste heat, and renewable heat sources such as solar thermal or fuel cells are potential application areas of the S-$CO_2$ cycle. In this paper, the current development progress of the S-$CO_2$ cycle is introduced. Moreover, a quick comparison of various S-$CO_2$ layouts is presented in terms of cycle performance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.219-230
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• 2011

The nuclear fuel cycle (NFC) analysis is a study to set a NFC policy and to promote systematic researches by analyzing technologies and deriving requirements at each stage of a fuel cycle. System analysis techniques are utilized for comparative analysis and assessment of options on a considered system. In case that NFC is taken into consideration various methods of the system analysis techniques could be applied depending on the range of an interest. This study presented NFC analysis strategies for the development of a domestic advanced NFC and analysis techniques applicable to different phases of the analysis. Strategically, NFC analysis necessitates the linkage with technology analyses, domestic and international interests, and a national energy program. In this respect, a trade-off study is readily applicable since it includes various aspects on NFC as metrics and then analyzes the considered NFC options according to the derived metrics. In this study, the trade-off study was identified as a method for NFC analysis with the derived strategies and it was expected to be used for development of an advanced NFC. A technology readiness level (TRL) method and NFC simulation codes could be utilized to obtain the required metrics and data for assessment in the trade-off study. The methodologies would guide a direction of technology development by comparing and assessing technological, economical, environmental, and other aspects on the alternatives. Consequently, they would contribute for systematic development and deployment of an appropriate advanced NFC.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.37-46
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• 2010

The development of advanced nuclear fuel cycle(ANFC) technology is essential to meet the national mission for energy independence via a nuclear option in Korea. The action target is to develop environmentally friendly, cost-effective measures to reduce the burden of long term disposal. The proper scenarios regarding potential radionuclide release from a repository have been developed in this study based on the advanced korean Reference Disposal System(A-KRS). To predict safety for the various scenarios, a new assessment code based on the GoldSim software has also been developed. Deterministic analysis indicates an environmental benefit from the ANFC as long as the solid waster from the ANFC act as a proper barrier.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.