• Analytical Science and Technology
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• v.13 no.3
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• pp.291-296
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• 2000

The SIMFUEL which composition is similar to PWR nuclear spent fuels was dissolved with a acid digestion bomb. An analytical conditions of ICP-AES for the direct determination of molybdenum in the uranium matrices without separation process were investigated. Based on the effect of uranium on molybdenum intensity. the most optimum wavelengths of molybdenum were found to be 202.030 and 203.844 nm. However, the method of standard additions is applied to overcome the effects of changing background caused by analyzing the sample solutions containing high concentration of uranium and the standard calibration solutions. The relative error of two methods, direct and indirect measurements with cation exchange resin separation procedures, was less than 5%. Therefore it was possible for this procedure to directly measure molybdenum in uranium matrices without separation. And this method was also applied to the determination of several percent of molybdenum in a U-Mo alloy.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Transactions of the Korean hydrogen and new energy society
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• v.17 no.2
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• pp.218-226
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• 2006

It can be obtained from hydrocarbon and water, specially production of hydrogen from natural gas is most commercial and economical process among the hydrogen production methods, and has been used widely. However, conventional hydrogen production methods are dependent on fossil fuel such as natural gas and coal, and it may be faced with problems such as exhaustion of fossil fuels, production of greenhouse gas and increase of feedstock price. Thermochemical hydrogen production by nuclear energy has potential to efficiently produce large quantities of hydrogen without producing greenhouse gases. However, nuclear hydrogen must be economical comparing with conventional hydrogen production method. Therefore, hydrogen production cost was analyzed and estimated for nuclear hydrogen as well as conventional hydrogen production such as natural gas reforming and coal gasification in various range.

• Carbon letters
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• v.4 no.1
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• pp.36-39
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• 2003

Carbon and carbon-based materials are used in nuclear reactors and there has recently been growing interest to develop graphite and carbon based materials for high temperature nuclear and fusion reactors. Efforts are underway to develop high density carbon materials as well as amorphous isotropic carbon for the application in thermal reactors. There has been research on coated nuclear fuel for high temperature reactor and research and development on coated fuels are now focused on fuel particles with high endurance during normal lifetime of the reactor. Since graphite as a moderator as well as structural material in high temperature reactors is one of the most favored choices, it is now felt to develop high density isotropic graphite with suitable coating for safe application of carbon based materials even in oxidizing or water vapor environment. Carboncarbon composite materials compared to conventional graphite materials are now being looked into as the promising materials for the fusion reactor due their ability to have high thermal conductivity and high thermal shock resistance. This paper deals with the application of carbon materials on various nuclear reactors related issues and addresses the current need for focused research on novel carbon materials for future new generation nuclear reactors.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.9-20
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• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.877-887
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• 2022

A new reactor concept is described that directly couples a supercritical CO₂ (sCO₂) power cycle with a CO₂-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO₂ concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (~20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3725-3731
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• 2023

Thorium dioxide (ThO₂)-silicon carbide (SiC) composite fuel pellets were fabricated via the spark plasma-sintering (SPS) method to investigate the role of the addition of SiC in enhancing the thermal conductivity of ThO₂ fuel. SiC particles with an average size of 1㎛ in 10 and 15 vol% were used to manufacture the composite pellets. The changes in the composites' densification, microstructure and thermal conductivity were explored by comparing them with pure ThO₂ pellets. The structural and microstructural characterization of the composite pellets has revealed that SPS could manufacture high-quality composite pellets without having any reaction products or intermetallic. The density measurement by the Archimedes principles and the grain size from the electron back-scattered diffraction (EBSD) analysis has indicated that the composites have higher densities and smaller grain sizes than the pellets without SiC addition. Finally, thermal conductivity as a function of temperature has revealed that sintered ThO₂-SiC composites showed an increase of up to 56% in thermal conductivity compared to pristine ThO₂ pellets.

• Journal of Ocean Engineering and Technology
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• v.37 no.5
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• pp.205-214
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• 2023

As the regulations on greenhouse gas emissions at sea become strict, efforts are being made to minimize environmental pollutants emitted from fossil fuels used by ships. Considering the large sizes of ships in conjunction with securing stable supplies of environment-friendly energy, interest in nuclear energy to power ships has been increasing. In this study, the neutron irradiation that occurs during the nuclear reactor operation and its effect on the structural responses of the stiffened-plate structures are investigated. This is done by changing the material properties of DH36 steel according to the research findings on the neutron-irradiated steels and then performing the structural response analyses of the structures using analytical and finite-element numerical solutions. Results reveal the influence of neutron irradiation on the structural responses of the structures. It is shown that both the strength and stiffness of the structures are affected by the neutron-irradiation phenomenon as their maximum flexural stress and deflection are increased with the increase in the amount of neutron irradiation. This implies that strength and stiffness need to be considered in the design of ships equipped with marine nuclear reactors.

• Proceedings of KOSOMES biannual meeting
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• 2018.06a
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• pp.95-95
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• 2018

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2006.10a
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• pp.293-294
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• 2006