• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.79-80
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• 2004

• Advances in Energy Research
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• v.1 no.1
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• pp.13-33
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• 2013

The present work is an attempt to provide an overview, about the status of R&D and current trends in Hydrogen Production using High Temperature Reactors. Bibliographic references from the INIS database, the Science Direct database and the NTIS database were downloaded and analyzed. Ten year data on the subject, published between 2001 and 2010, was selected for the study. Appropriate qued ry formulations on these databases, resulted in the retrieval of 621 unique bibliographic records. Using the content analysis method, all the records were analyzed. Part One of the analysis details Scientometric R&D indicators, Part Two is a subject-based analysis, grouped under: A. International Initiatives and Programmes for Hydrogen Production; B. European R&D initiatives for Hydrogen production; C. National Initiatives and Programmes for Nuclear Hydrogen Production; D. Reactor Technologies for Nuclear Hydrogen production; E. Fuel Developments; F. Hydrogen Production Processes using HTRs and G. Materials Consideration for Nuclear Hydrogen Production. The results of this analysis are summarized in the study.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• Journal of the Semiconductor & Display Technology
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• v.14 no.2
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• pp.61-65
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• 2015

SiC composite materials are usually used to very high temperature condition such as thermal protection system materials at space vehicles, combustion chambers or engine nozzles because they have high specific strength and good thermal properties at high temperature. One of the most widely used fabrication methods of SiC composites is the chemical vapor infiltration (CVI) process. During the process, chemical gases including Si are introduced into porous preform which is made by carbon fibers for infiltration. Since the processes take a very long time, it is important to reduce the process time in designing the reactors and processes. In this study, both the gas flow and heat transfer in the reactors during the processes are analyzed using a computational fluid dynamics method in order to design reactors and processes for uniform, high quality SiC composites. Effects of flow rate and heater temperature as process parameters to the infiltration process were examined.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.06a
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• pp.199-200
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• 2009

• Journal of Environmental Health Sciences
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• v.29 no.1
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• pp.43-50
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• 2003

This study was performed to compare the effects of rice straw and towage sludge cake as bulking materials on temperature, pH, weight and volume reduction, porosity, C/N ratio, salinity, and conductivity in aerobic composting of food wastes. Volume ratios of food wastes to rice straw in reactor control, RS-1, RS-2, RS-3 and RS-4 were 4:0, 4:1, 4:2, 4:3 and 4:4, respectively. Weight ratios of food wastes to sewage sludge rake in reactor control, SL-1, SL-2, SL-3 and SL-4 were 4:0, 4:1, 4:2, 4:3 and 4:4, respectively. Reactors were operated for 24 days with 1 hour stirring by 1 rpm and 2 hours aeration per day. The values of pH of food waters, rice straw and sewage sludge cake were 4.39, 7.40 and 5.79, respectively. The lowering of the volume ratio of food wastes to rice straw resulted in the high reaction temperature and the fast weight and volume reduction rates. The lowering of the weight ratio of food wastes to sewage sludge cake resulted in the slow weight and volume reduction rates. C/N ratio in control was larger than that in rice straw containing reactors, and that in rice straw containing reactors was larger than that in sewage sludge cake containing reactors. Salinity and conductivity in reactors were condensed and increased by reaction days.

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.365-374
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• 2001

In order to examine the in-reactor performance of very-high-density dispersion fuels for high flux performance research reactors, U-l0wt.%Mo microplates containing centrifugally atomized powder were irradiated at low temperature. The U-l0wt.%Mo dispersion fuels show stable in- reactor irradiation behaviors even at high burn-up, similar to U$_3$Si$_2$ dispersion fuels. The atomized U-l0wt.%Mo fuel particles have a fine and a relatively uniform fission gas bubble size distribution. Moreover, only one of third of the area of the atomized fuel cross-sections at 70a1.% burn-up shows fission gas bubble-free zones, This appears to be the result of segregation into high Mo and low Mo.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.115.1-115.1
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• 2003