[en] According to the trend toward high burn-up nuclear fuel, it is issued recently that whether current LOCA criteria are also applicable to high burn-up nuclear fuel cladding or not. The real behaviors of high burn-up nuclear fuel cladding under LOCA are estimated and the main issues are analysed as follows. Whether oxidation during the normal operation should be inclusive in ECR calculation or not. The hydrogen effects on ductility of cladding on quenching. Applicability of Niobium-containing high burn-up nuclear fuel cladding to urrent LOCA criteria. Acceleration of steam oxidation kinetics under high pressure steam

[en] Spent fuel inventories to be disposed of finally and design base spent fuel were determined. Technical and safety criteria for a geological repository system in Korea were established. Based on the properties of spent PWR and CANDU fuels, seven repository alternatives were developed and the most promising repository option was selected by the pair-wise comparison method from the technology point of view. With this option preliminary conceptual design studies were carried out. Several module, e.g., gap module, congruent release module were developed for the overall assessment code MASCOT-K. The prominent overseas databases such as OECD/NEA FEP list were are fully reviewed and then screened to identify the feasible ones to reflect the Korean geo-hydrological conditions. In addition to this the well known scenario development methods such as PID, RES were reviewed. To confirm the radiological safety of the proposed KAERI repository concept the preliminary PA was pursued. Thermo-hydro-mechanical analysis for the near field of repository were performed to verify thermal and mechanical stability for KAERI repository system. The requirements of buffer material were analyzed, and based on the results, the quantitative functional criteria for buffer material were established. The hydraulic and swelling property, mechanical properties, and thermal conductivity, the organic carbon content, and the evolution of pore water chemistry were investigated. Based on the results, the candidate buffer material was selected

[en] Design, high-voltage test, and installation of 6 MW ICRF heating system for KSTAR is completed. The antenna demonstrated satisfactory standoff at high voltages up to 41 kV for 300 sec. The result indicates good power handling capabilities of the antenna as high as 10 MW/m2. This power density is equivalent to RF power coupling of 6 MW into a 4 Ω/m target plasma, and is typical of advanced tokamak heating scenarios. In addition, vacuum feed through, DC break, and liquid stub developed for 300 sec operation are installed, as well as a 2 MW, 30-60MHz transmitter. The transmitter successfully produced output powers of 600 kW continuously, 1.5∼1.8 MW for 300 sec, and 2 MW for 100 msec or shorter pulses. A realtime control system based on DSP and EPICS is developed, installed, and tested on the ICRF system. Initial results from feasibility study indicate that the present antenna and the transmission lines could allow load-resilient operation on KSTAR. Until the KSTAR tokamak start to produce plasmas in 2008, however, hands-on operational experiences are obtained from participating in ICRF heating experiments at ASDEX and DIII-D tokamaks arranged through international cooperation

[en] Recently, international co-works on failure behavior of high burn-up nuclear fuels have been briskly progressing in research reactors. In this paper, the test results of RIA-simulating test reactors such as PBF, SPERT, IGR, BIGR, CABRI, NSRR are introduced

[en] The final objectives of this project are as follows; Development of the key techniques for the ITER blanket first wall - ITER blanket first wall design and developing related techniques - Blanket first wall qualification test and verification of manufacturing technology - Qualification of the bonding technique and development of the fabrication method for the ITER blanket first wall. We are technically surveying the ITER system design data, the insufficient part of ITER design, and required R and D items and so on. And we are developing the manufacturing process and joining technology for the ITER blanket to complete the procurement with qualified blankets system. To evaluate the soundness of manufacturing process, specimen and mock-up tests are being prepared. Finally, we can obtain the key technology of nuclear fusion reactor especially on the blanket design, joining and manufacturing technology through the present project and these technologies will help the construction of Korea fusion Demo reactor and the development of commercial nuclear fusion reactor in Korea

[en] The final objectives of this project are as follows; Development of the key techniques for the liquid type TBM for ITER: Developing plan for leading and participating liquid TBM concepts; Estimating cost and schedule according to development schedule and managing technologies; Developing integrated design system and completing the engineering design for liquid TBM; Developing the key technologies for the liquid TBM; Construction of performance test systems for liquid TBM and verification of the performance. We are technically surveying the ITER system design data, the insufficient part of ITER design, and required R and D items and so on. In Korea, HCML TBM, liquid type breeder with lithium or lead lithium, has been studied during the past years to develop a tritium breeding technology for tritium self-sufficiency of nuclear fusion reactor and the TBM was proposed to be tested in ITER. In this study, we can obtain the key technology of nuclear fusion reactor especially on the TBM design, analysis and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea

[en] Although it is expected that the decommissioning of a nuclear power plant will happen since 2020, the need of partial decommissioning and decontamination for periodic inspection and life extension has been on an increasing trend and domestic market has gradually been extended. Therefore, in this project the decommissioning DB system on the KRR-1 and 2 was developed as establishing the information classification system of the research reactor dismantling and the structural design and optimization of the decommissioning DB system. Also in order to secure the reliability and safety about the dismantling process, the main dismantling simulation technology that can verify the dismantling process before their real dismantling work was developed. And also the underwater cutting equipment was developed to remove these stainless steel parts highly activated from the RSR. First, the its key technologies were developed and then the design, making, and capability analysis were performed. Finally the actual proof was achieved for applying the dismantling site. an automatic surface contamination measuring equipment was developed in order to get the sample automatically and measure the radiation/radioactivity

[en] In the decommissioning planning stage, it is important to select the optimized decommissioning process considering the cost and safety. Especially the selection of the optimized decommissioning process is necessary because it affects to improve worker's safety and decommissioning work efficiency. The decommissioning process evaluation technology can provide the optimized decommissioning process as constructing various decommissioning scenarios and it can help to prevent the potential accidents as delivering the exact work procedures to workers and to help workers to perform decommissioning work skillfully. It's necessary to measure the radioactive contamination in the highly contaminated facilities such as hot-cells or glove-boxes to be decommissioned for decommissioning planning. These facilities are very high radiation level, so it is difficult to approach. In this case the detector system is preferable to separate the sensor and electronics, which have to locate in the facility outside to avoid the electric noise and worker's radiation exposure. In this project, we developed the remote detection system for radiation measurement and signal transmission in the high radiation area. In order to minimize worker's exposure when decommissioning highly activated nuclear facilities, it is necessary to develop the remote handling tool to perform the dismantling work remotely. Especially, since cutting, measuring, and decontamination works should be performed remotely in the highly activated area, the remote handling tool for conducting these works should be developed. Therefore, the multi-purpose dismantling machine that can measuring dose, facility cutting, and remote handling for maintenance and decommissioning of highly activated facility should be needed

No abstract available

[en] Recently, international co-works on failure behavior of high burn-up nuclear fuels have been briskly progressing in research reactors. In this paper, the test results of CABRI and NSRR test programs focused on the mechanical properties of fuel cladding during the RIA(Reactivity Initiated Accident) conditions are introduced. According to the results from CABRI and NSRR test program up to now, the primary failure mechanism was clarified to be hydride-assisted PCMI(Pellet Cladding Mechanical Interaction). In addition, it turns out that the oxide spallation and local hydride blister of cladding play deleterious roles in the RIA failure