[en] A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

[en] Five-hole pitot tube is an effective detector that could measure a three dimensional average flow field on a complex geometry. At the present study, have been mainly used in the field of aerodynamics and nautics, the five-hole pitot tube is extensively investigated to apply on the nuclear engineering. Five-hole pitot tube could measure the three dimensional velocity to make use of a relationship between pressure energy and kinetic energy from Bernoulli's equation; therefore, the report shortly overviewed the definition, units, and transducers of pressure and then detaily was described about the pitot tube. For five-hole pitot tube, history, kinds and fabrication methods were briefly provided. The calibration methods for the five-hole pitot tube were deeply introduced in various methods according to simple concept but complex process. Additionally, causeses of detection errors and estimation of uncertainty were included in the present report. Optical measurement and how wire anemometers are difficult to detect the flow velocity under environmental such as tight lattice bundle geometry, dusty flow and high temperature fluid. One of alternatives to overcome the diffculty is the five-hole pitot tube

[en] A subchannel code system is developed for the thermal-hydraulic analysis of SMART core, and the applicability and accuracy of the code is assessed for various experimental data with rod bundles. MATRA is a subchannel analysis code calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. MATRA has been provided with an improved structure and code functions to give more convenient user environment. Improvement of various models enhances the convergence and accuracy of the code: those include the numerical solution scheme for the crossflow, the void fraction model, and the lateral transport model, and so on. A turbulent mixing model considering void drift phenomenon is devised by employing the two-phase mixing test data under PWR and BWR conditions. MATRA/SR-1 CHF correlation system is developed from local conditions of rod bundle CHF data calculated by MATRA. The optimized 1/8 core lumping models are developed for the analysis of the thermal margins of SMART core at steady-state and transient conditions

[en] This report was written as following contents, to provide experience of work place and experiments using HANARO and its facilities, to provide a chance for that selection of various major scopes in the future for university students of science and technology by KAERI. Purpose of this research is to offer a specialized education opportunity by using HANARO and its facilities to university students by developing and operating various curriculum for future users. This is purposeful in various practical ways and achieves follow -up research for this area. First, this practice offers the opportunities to university students by developing various and continuously operating research processes by using HANARO. Second, reactor experiments of the university students contributes to the training for specialist as the training on operating reactor practices are continuous. Third, student experiments for the university students of science and technology are purposeful in developing and magnifying base-users as well as the related specialists of the nuclear power industry hereafter. Finally, training courses utilizing nuclear reactor facilities activate and expand various fields, and they become to important resources for establishing a nuclear energy policy and technology

[en] Nuclear Training Center (NTC) has concentrated its efforts on the systemization and specialization of education and training programs and has actively carried out diverse activities to create new nuclear courses based on its experience accumulated over the past years. The systematic and comprehensive education systems have been set up by streamlining the education systems for internal employees conducted sporadically over the past years. For the development of efficient education courses, expansion and diversification of education and training programs have been carried out based on the study of the Systematic Approach to Training (SAT) methodology and a survey of manpower development in on-site industry. Furthermore, despite the decrease in the number of potential trainees due to various government-enforced regulations and the increase in the number of other nuclear education institutions, the consistent efforts to develop new courses and to improve previous programs based on evaluations from the trainees have led to an upgrade in the level of education and the efficiency of the operation compared to previous years. In terms of development, 7 education programs were improved accordingly and 10 new courses were created. The year 2008 was especially considered a landmark, as NTC was authenticated as one of the best Human Resource Development (HRD) institutions for providing various kinds of education in the field of nuclear energy. Today, it is recognized as the leading HRD center in Korea

[en] A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

[en] The information As application of digital technologies being extended to the safety related system in a nuclear power plant, NRC of United States has issued recently guides and instructions which state verification and validation (V and V) of man-machine interface system (MMIS) should be performed using a dynamic simulator. For V and V of Korean Next Generation Reactor's MMIS, this paper proposes a configuration of an engineering simulator based on Modular Modeling System (MMS), which has been world-widely used and its performance has been verified. (author)

[en] The main objective of the present study is to perform a comparative study of five existing correlations that have been selected and identify the best performing correlations in the subchannel pressure drop analysis of a wire-wrapped fuel assembly by means of directly comparing with experimental data obtained in the present work. For this purpose, a series of water experiments have been performed using a helical wire-wrapped 19-pin fuel assembly for various combinations of test parameters. Four different test sections that have different pitch to rod diameter ratios (P/D) and wire lead length to rod diameter ratios (H/D) have been fabricated. A series of pressure drop measurements were made to obtain friction factors for these four test sections. A total of 293 data were obtained and the present along with existing data are used in the present comparative study of existing correlations. The results of this study show that both the original and the simplified Cheng and Todreas correlations give the best agreement with experimental data for all flow regions

[en] This report was written as following contents, to develop a program for university students majoring in science and technology, which is intended to provide the students with opportunities to obtain hands on experience and knowledge on various nuclear technology, through experiments using HANARO and its facilities. Thus obtain experience and knowledge are expected to be a great help for their current study and for their selection of a specific future study area. The purpose of this research is as follows: development of various curricula for specific research using HANARO and continuous operation of the developed curricula to provided university students with opportunities to use HANARO as part of their university study, continuous operation of research reactor experimental programs for university students in nuclear field to make contribution to cultivating specialists, development and operation of training programs of experiments using research reactor for university students majoring in nuclear engineering and also for university students majoring in diverse fields of science and technology such as physics, advanced metallurgy, mechanical engineering, energy engineering, radiological science, nanoscience, etc. to cultivate future potential users of HANARO as well as broadening the user group. As a whole, 451 students from 19 universities have completed the courses of the programs developed and offered by this project. Also, 6 textbooks have been developed to support the programs

[en] In this paper the method to evaluate shielding deficiency by gamma scanning test was presented and verified theoretically by Monte Carlo code which is one of the best effective method for radiation shielding calculation. The cylindrical shielding model was selected to evaluate shielding deficiency by gamma scanning test. First, the reference shielding according to the design requirement of cask was fabricated specially and reference values were measured with Co-60 source and scintillation detector. As a result with which calculated the reference values, it is shown that maximum deficiency thickness for lead of true cylindrical shielding model was 12mm. To verify this, thickness of lead was calculated by MCNP code and maximum deficiency thickness was 11.6mm. The experimental result obtained by the use of reference shielding was in good agreement with the theoretical result within 4.1%. So, this method can be applied to inspect the shielding ability for great shielding or cask which the radioactive material is used. To perform measurement more exactly, the further work on the development of measuring equipment to display the results on the screen will be required