[en] The research scope reported herein consists of the nuclear data service, evaluation of nuclear data, processing of evaluated data, and measurements of cross-sections. The web service provides the table of nuclides, evaluated nuclear data libraries, experiments data, and nuclear data for medical application. In addition, a domestic nuclear data network has been organized to exchange relevant information between producers and users of nuclear data. The nuclear data evaluated during the project period include neutron cross-sections of 20 fission products and 5 Dy isotopes, photo-nuclear reaction data of 143 nuclides, charged-particle cross-sections of 61 nuclides for medical and industrial applications, and high-energy data of 9 nuclides including Al-27. Several computer code systems for the evaluation has been imported, enhanced, or developed. The NJOY code system has been settled down for the processing of evaluated data. WIMS/D, CASMO, and HELIOS libraries for the reactor core nuclear design and KASHIL-E6 library for shielding analyses are the major group-constant libraries produced, benchmarked, and provided. A processing methodology has been enhanced for cross-sections at extremely low and high energy region. A test time-of-flight facility with a 100 MeV electron linac and a facility using the activation method with a VDG accelerator have been built to measure neutron cross-sections in the energy region below 100 keV and in the region of 1∼2 MeV, respectively. The neutron cross-sections of Cu etc. were measured at these facilities. Meanwhile, the capture cross-sections of Dy, Hf, and Th-232 were measured at foreign facilities

[en] An internet portal site which can be a hub web site of information, was developed and disclosed to enhance the information production and dissimination. The representative functions of the portal site are to provide a site map which provide categorized service of domestic and global internet home pages, and a bulletin board, a closed user group. The serviced databases consist of an encyclopedia of nuclear information, which contains detailed description of nuclear relevant topics, a software database for computer program services, a newspaper database for proving up to date news articles, a descriptive database on the incident and accident on the nuclear power plant, and a central search engine on the bibliographic informations. The training lecture texts for the experts on the field of nuclear energy and radiation technology was converted into HTML formatted text and are on service on an internet web site. Six coursewares for the students studying the nuclear engineering are developed and six digital education platforms are introduced for the nuclear engineering departments of domestic universities. The platforms are used in teaching students utilizing the courseware intergrated with engineering simulation programs for substituting experiments which are difficult, dangerous and sometimes impossible in real situation

[en] The suitable turbulence model is found to be required in the course of establishing a proper analysis methodology for thermal stripping phenomena. Two different turbulence models of κ-ε model and Full Reynolds Stress (FRS) model, are applied to analyze unsteady turbulent flows with temperature variation. Test case is cross jet sodium flow in pipe with T-junction. Analysis yields the conclusion that 3-D computation with turbulence model is generally able to predict physical characteristics of flow. FRS predicts fast diffusion transport effect. However, modified modeling for near wall effect is required to improve its heat transfer characteristic analysis

[en] LES is applied to analyze unsteady turbulent triple-jet water flows which are commonly found in thermal stripping phenomena. 3-D computations are performed with LES of Smagorinsky-Lilly SGS model. The LES successfully produces time history of turbulence variables which could be used to evaluate magnitude and frequency of fluctuating temperature. Comparisons of the LES results with experimental data show that the LES underestimates mixing activity in flow field. Further study is required to find fine grid system effect and inlet turbulence intensity effect

[en] This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

[en] LES is applied to analyze transient turbulent flows which may cause thermal stripping. The LES is formulated based on Sub-Grid k model. For evaluating performance of the LES, two test cases of vertical water jet flow and parallel sodium jet flow are selected. Through the analysis, the LES confirms that it has proper ability to predict time dependent flow variables such as magnitude of temeprature fluctuation and its frequency. For the better prediction, however, the LES is required to be improved through further evaluation of various SGS models and implementation of temperature SGS model into its energy equation

[en] To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

[en] In analysis of unsteady turbulent flows with temperature variation, 3 different turbulence models of κ-ε model, modified κ-ε model, and Full Reynolds Stress(FRS) model, are applied. 3 Test Case are selected for verification. These are vertical jet flows with water and sodium, and parallel jet flow sodium. For overall verification of turbulence models, test cases are analyzed with 2-D and 3-D assumptions. Analysis yields the conclusion that 3-D computation with FRS betters others. However, modified modeling for near wall effect is required to improve its heat transfer characteristics analysis

[en] Enhancement of public understanding on the nuclear safety charter - Framing draft of preamble and articles in nuclear safety charter - Collection of professionals' concerns in the field of nuclear - Completion of explanatory text on the preambles and articles in charter. Enhancement of public confidence on the nuclear safety - Setting new publicity strategy involving animation technic to improve harmony with public - Development of public information prototype and materials about preamble and articles in the nuclear safety

[en] The neutron-induced transmutation is one of the most promising methods of reducing the radio-toxicity of nuclear spent fuel for the long-term HWR management aspect. In this research, we investigate the effectiveness of a molten-salt reactor for transmutation. A molten salt reactor can operate with continuous feeding of and removing of the FPs. Basic approach for developing design characteristics adopt the AMBIDEXTER concept which had been conceptually designed at Ajou Univ. The ORIGEN2-HELIOS-SQUID code system was developed for the lattice and core design analyses. As the reference fuel material concentration of Pu and Minor Actinides were chosen from 33000MWD/teU PWR spent fuel. Because NaF-ZrF₄ based fuel has harder neutron spectrum than LiF-BeF₂, that was selected as bare material of fuel salt. Results of the simulation, demonstrate that 250MW AMBIDEXTER-based transmutation reactor could remove 257g/day of actinide. We also computed transient of TRU mass fraction, removal and product rates