• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.292-297
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• 1997

An alternating conditional expectation (ACE) algorithm, a kind of non-parametric regression method, is proposed to generate empirical correlations automatically. The ACE algorithm yields an optimal relationship between a dependent variable and multiple independent variables without any preprocessing and initial assumption on the functional forms. This algorithm is applied to a collection of 12,879 CHF data points for forced convective boiling hi vertical tubes to develop a new critical heat flux (CHF) correlation. The meat root mean square, and maximum errors of our new correlation are -0.558%, 12.5%, and 122.6%, respectively. Our CHF correlation represents the entire set of CHF data with an overall accuracy equivalent to or better than that of three existing correlations.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• The Transactions of the Korean Institute of Electrical Engineers P
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• v.58 no.2
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• pp.227-230
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• 2009

In this paper, we performed the optimum design of reactors for matrix-type superconducting-fault current limiters (SFCLs), using electromagnetic analysis tools. We decided a optimun position within a reactor for superconducting elements of current-limiting parts and trigger parts from the calculation of magnetic flux internsity for reactor structures. Also we decided effective distance length between two reactors through the analysis of the distribution of magnetic field, according to distance lengths between them. We designed and characterized matrix-type SFCLs, based on our optimum design of a reactor. We confirmed uniform distribution of a fault current, resulted from the improvement of simultaneous quench characteristics within our matrix-type SFCL.

• Journal of Electrical Engineering and Technology
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• v.2 no.3
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• pp.289-293
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• 2007

The fault current limiting characteristics of the separated and the integrated three-phase flux-lock type superconducting fault current limiters (SFCLs) were analyzed. The three-phase flux-lock type SFCL consisted of three flux-lock reactors and three $high-T_c$ superconducting (HTSC) elements. In the integrated three-phase flux-lock type SFCL, three flux-lock reactors are connected on the same iron core. On the other hand, three flux-lock reactors of the separated three-phase flux-lock type SFCL are connected on three separated iron cores. The integrated three-phase flux-lock type SFCL showed the different fault current limiting characteristics from the separated three-phase flux-lock type SFCL that the fault phase could affect the sound phase, which resulted in quench of the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the separated and the integrated three-phase flux-lock type SFCLs according to the ground fault types were compared.

• 한국전산유체공학회:학술대회논문집
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• 2005.10a
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• pp.31-34
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• 2005

This article has investigated the spatial distribution of the solid particles in polymerization reactors using CFD analysis (FLUENT v. 6.2.1). The suspension of the solids in stirred reactors is affected by a number of parameters including particle diameter, vessel shape, impeller size, impeller speed, and rotating direction of stirrer. The degree of solids suspension in the vessel was quantified with a statistical average value, ${\sigma}^2$. The best stirring conditions were determined based on ${\sigma}^2$, which was found to depend on the vessel bottom shape.

• 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.

• 제어로봇시스템학회:학술대회논문집
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• 1993.10a
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• pp.595-600
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• 1993

The heat of reaction has been estimated from heat balance relationships around the reactor. The heat balance equations were formulated with the assumptions that the reactor temperature is uniformly distributed and the jacket temperatures are axially distributed. We have obtained the temperature distribution of jacket contents by FDM. And then, we have rearranged the heat balance equations so that the heat of reaction can be estimated from the finite number of temperature measurements, i.e., temperatures of the reactor contents, at the jacket inlet and outlet, respectively. The proposed method for reaction heat estimation on were applied to industrial batch reactors ; one is ABS polymerization reactor and the other is SAN polymerization reactor. We have also examined the variation of overall heat transfer coefficients for the reactors during reaction.

• Journal of Environmental Science International
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• v.13 no.1
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• pp.69-77
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• 2004

In this study, sulfate reduction reaction was used to increase the decomposition of organics, which is the most critical factor for the stabilization of a landfill site. Composite of sewage sludge, papers, and incineration ashes was used in the column. The experimental results indicated that out of 10 reactors, the reactors 3, 4, 8. and 9 showed higher organics (i.e., TOC) removal rate than that in the absence of sulfate. The organics removal rates (K) in R3 and R9 were 8.65e$\^$-4/d and 3.82e$\^$-4//d, respectively. The times to reach 10％ of initial concentrations in R3 and R9 was 7.3 and 16.5 years, respectively, showing faster organics decomposition rates in these reactors.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.177-198
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• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.