[en] Non-destructive test system was installed at hot-cell(M1) in IMEF(Irradiated Materials Examination Facility) more than 14 years ago for the diametric measurement of fuel rod. But this system must be needed to be remodeled for the effective operations. In 2006, the system was upgraded for 2 months. Some of electronic parts were added in PLC panel, and operating panel was re-designed for the remote control. To operate the fuel bench by computer, AD converter and some I/O cards were installed in computer. All of software were developed in Windows-XP system instead of DOS system. Control programs were made by visual-C language. After upgrade of system, operator can control diametric measurement easily. Series of tests were carried out automatically and data of each test were saved continuously. With consideration of ECT(Eddy Current Test) installation, the computer program and hardware were set up as well. But ECT is not installed yet, so we have to check abnormal situation of program and hardware system. It is planned to install ECT in 2007

[en] Swelling of U₃Si and U₃Si₂ is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs

[en] The chemical interaction between metallic fuel and cladding is important in designing the fuel pin of the KALIMER. When metal fuel and cladding are contacted, the elements in fuel and cladding are inter-diffuse each other, forming the reaction layers at interface. The reaction layers may cause two important factors in aspects of fuel pin integrity. Firstly, it degrades cladding strength by reducing effective cladding thickness. Secondly, these layers accelerate eutectic reaction at transient conditions. To evaluate these phenomena, the diffusion couple experiment is planned by using metal fuels with various zirconium contents and HT-9 steel. The U-Zr fuel alloys will be used for the experiment with the different zirconium contents, these are 8, 10 and 12 weight %. This experiment aims to evaluate the effects of zirconium content on the chemical reaction. Furthermore, the reaction rate and threshold temperature of the eutectic melting will be determined as a function of the zirconium content. This document describes the detail experimental specifications for the eutectic reaction such as test setup, test requirements and test procedure. (author). 10 refs

[en] In order to determine design allowable stress intensity, the creep work(W_c) and time(t) equation defined as W_c t^p=B was verified using experimental results. For this purpose, the creep tests for generating 1% strain in 10⁵h for commercial type 316 stainless steel were conducted with different stresses; 160 MPa, 150 MPa, 145 MPa, and 140 MPa at 593 .deg. C. Results between log W_c and log t showed a linear relation well, and the allowable stress intensity value of the W_c - t equation showed good agreement to that of Isochronous Stress-Strain Curves(ISSC) presented on ASME-NH code. The equation can be simply obtained with only several short-term 1% strain data without ISSC

[en] The project has been carried out for 2 years in stage III in order to achieve the final goals of performance verification of the developed materials, after successful development of the advanced high temperature material technologies for 3 years in Stage II. The mechanical and thermal properties of the advanced materials, which were developed during Stage II, were evaluated at high temperatures, and the modification of the advanced materials were performed. Moreover, a database management system was established using user-friendly knowledge-base scheme to complete the integrated-information material database in KAERI material division

[en] Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

[en] Data for metal fuel rod and fuel assembly performance code is collected, fuel rod fabrication technologies for engineering scale metal fuel fabrication facility is developed, and conceptual design for UFMF construction is performed to obtain the specific design approval from the regulatory authority planned in 2020. Mechanical tests such as creep and transient tests are performed on HT9 and FC92 cladding tubes. In-pile creep and neutron swelling data up to 39dpa irradiation rate are obtained. The technologies for property tests such as density, alloy element and impurity evaluation, and microstructures are performed in addition to various thermal and mechanical properties for property assurance in future engineering metal fuel rod fabrication. Engineering scale metal fuel slug casting is developed, fuel rod fabrication and assembly device are designed, and FC92 cladding tube is fabricated by industry (Iljin steel manufacture company) to develop U-Zr metal fuel mass production technologies. Metal scrap recycle is researched to recycle uranium resources and enhance metal fuel economy. Fuel quality assurance and preliminary property criteria are established and property inspection items were produced (fuel rod fabrication, sodium melting/bonding, end plug welding procedure) to establish UFMF metal fuel preliminary management system. Construction, engineering, machinery, pipe arrangement, and electricity design criteria was prepared. Organization and conception of physical protection system and facility safety management, entrance management system were actualized. Moreover, the economic analysis of the facility was performed

[en] Proposed materials, Mod.9Cr-1Mo steel (32 mm thickness) and 9Cr-1Mo-1W (100 mm thickness), for the reactor vessel were procured, and welded by the qualified welding technologies. Welding soundness was conformed by NDT, and mechanical testings were done along to weld depth. Two new irradiation capsules for use in the OR test hole of HANARO were designed and fabricated. specimens was irradiated in the OR5 test hole of HANARO with a 30MW thermal power at 390±10 .deg. C up to a fast neutron fluence of 4.4x10¹⁹ (n/cm²) (E>1.0 MeV). The dpa was evaluated to be 0.034∼0.07. Base metals and weldments of both Mod.9Cr-1Mo and 9Cr-1Mo-1W steels were tested tensile and impact properties in order to evaluate the irradiation hardening effects due to neutron irradiation. DBTT of base metal and weldment of Mod.9Cr-1Mo steel were -16 .deg. C and 1 .deg. C, respectively. After neutron irradiation, DBTT of weldment of Mod.9Cr-1Mo steel increased to 25 . deg. C. Alloy 617 and several nickel-base superalloys were studied to evaluate high temperature degradation mechanisms. Helium loop was developed to evaluate the oxidation behaviors of materials in the VHTR environments. In addition, creep behaviors in air and He environments were compared, and oxidation layers formed outer surfaces were measured as a function of applied stress and these results were investigated to the creep life

[en] The study is to characterize the structural materials under the high temperature, one of the most significant environmental factors in nuclear systems. And advanced materials are developed for high temperature and/or low activation in neutron irradiation. Tensile, fatigue and creep properties have been carried out at high temperature to evaluate the mechanical degradation. Irradiation tests were performed using the HANARO. The optimum chemical composition and heat treatment condition were determined for nuclear grade 316NG stainless steel. Nitrogen, aluminum, and tungsten were added for increasing the creep rupture strength of FMS steel. The new heat treatment method was developed to form more stable precipitates. By applying the novel whiskering process, high density SiC/SiC composites with relative density above 90% could be obtained even in a shorter processing time than the conventional CVI process. Material integrated databases are established using data sheets. The databases of 6 kinds of material properties are accessible through the home page of KAERI material division

[en] This paper reviewed systematically a state-of-art of irradiation creep for stainless steels to provide a background information for performing irradiation creep tests and establishing the creep model for advanced domestic steels effectively. An irradiation creep model of SFR core materials is necessary to apply to the fuel cladding and assembly materials of domestic SFR reactor system. The document of in-reactor irradiation creep has been obtained by investing a long time and large-scale cost using limited experimental research reactors. This paper will provide the knowledge to understand the irradiation creep and to obtain the background information of advanced domestic steels, so that it hopes to practically apply for timely producing the documents of irradiation creep of advanced domestic steels necessary for the national SFR program