[en] The newly established organization KRMC (Korea radioactive waste management corporation) which is responsible for all kinds of radioactive waste generated in the Republic of Korea launched the PWR spent fuel dry storage research project in June 2009. This project has objectives to develop a storage system and evaluate the integrity of PWR fuel in dry storage. The project consists of three steps. At first step, it would develop own degradation models by referring to pre-exist good models and develop the hot test scenarios. Second step, test facilities would be constructed and used for testing the degradation behaviour in each mechanisms and in total. As a final step, total evaluation code would be developed by integrating each degradation model produced in the first step and the test data produced in the second step. All the activities would be summarized into a report and applied to licensing work. The Republic of Korea PWR spent fuels have unique characteristics of various fuel types (array type, clad material) and high capacity factor (maximum usage of fuel which is bad for integrity). These facts could impact on the research ranges of experimental data needed for degradation evaluation. In this research, spent fuel performance data concerning long term dry storage will be analysed and the major degradation mechanisms like creep and hydride behaviour will be studied and proposed for Korean PWR spent fuels

[en] This report contains an investigation of the disposal concept, the current state of technology, and safety and technical criteria relevant to geological HLW disposal system developed so far by the U.S., Canada and Sweden. At present, for the design criteria and license requirements of the repository system the U.S. is considering even the substantial technical specification, while other countries have determined a tentative radiological safety goal as a design target. For the reference disposal concepts, the U.S.'s system has been developed at bases on the horizontal emplacement method of the waste package (without buffer surrounding waste package) in disposal tunnel horizontally bored in 300 m - depth of saturated crystalline rock mass. Such disposal concepts of three nations, in principle, are based at the retrievable option of spent fuel or HLW disposed of when the political and safety problems would be issued during the post-closure phase. For the thermomechanical safety of the underground repository system, the U.S limits the surface cladding temperature of spent fuel <350 deg C wall temperature of disposal tunnel < 200 deg C as the most thermal constraints, while Canada and Sweden limit the surface temperature of waste container and the max. temperature of buffer material < 80 ∼ 100 deg C. (author). 26 refs., 10 tabs., 62 figs

[en] A database program for analyzing the characteristics of spent fuels was developed, and A-SOURCE program for characterizing the source term of HLW from advanced fuel cycles. A new technique for developing a copper canister by introducing a cold spray technique was developed, which could reduce the amount of copper. Also, to enhance the performance of A-KRS, two kinds of properties, thermal performance and iodine adsorption, were studied successfully. A complex geological disposal system which can accommodate all the HLW (CANDU and HANARO spent fuels, HLW from pyro-processing of PWR spent fuels, decommissioning wastes) was developed, and a conceptual design was carried out. Operational safety assessment system was constructed for the long-term management of A-KRS. Three representative accidental cases were analyzed, and the probabilistic safety assessment was adopted as a methodology for the safety evaluation of A-KRS operation. A national program was proposed to support the HLW national policy on the HLW management. A roadmap for HLW management was proposed based on the optimum timing of disposal

[en] For the purpose of developing the geological disposal system proposed in this research project successfully, the analysis of source terms of various waste forms, conceptual design and performance assessment of the engineered barrier system including a disposal canister, and the cost analysis based on the domestic unit cost should be carried out. To this end the following research items will be developed at this stage: Ο Development of a technology for assessing the source-term of advanced fuel cycle - Verification and enhancement of a source-term assessment program - Radiological performance assessment of A-Kfs - Assessment of national inventories of major radionuclides Ο Development of disposal canisters for HLA from pyro-processing - Development of canister sealing technology - Estimation of corrosion life-time of canisters with Korean Ebs - Analysis of shear stress on the canister with Korean Ebs Ο Development of a geological disposal system for HLA from advanced fuel cycle - Hydrological analysis for determining the layout of disposal tunnels - Preparation of database regarding the domestic unit disposal cost - Thermal analysis of disposal tunnels with ventilation during operation period - Feasibility assessment of A-Kfs from the technical, safety, and economical points of views

[en] A KRS is suggested through design requirement analysis of the buffer and the canister which are the constituent of disposal system engineered barrier and HLW management plans are proposed. In the aspect of radionuclide retention capacity, the thickness of the buffer is determined 0.5m, the shape to be disc and ring and the dry density to be 1.6 g/cm³. The maximum temperature of the buffer is below 100 .deg. which meets the design requirement. And bentonite blocks with 5 wt% of graphite showed more than 1.0 W/mK of thermal conductivity without the addition of sand. The result of the thermal analysis for proposed double-layered buffer shows that decrease of 7 .deg. C in maximum temperature of the buffer. For the disposal canister, the copper for the outer shell material and cast iron for the inner structure material is recommended considering the results analyzed in terms of performance of the canisters and manufacturability and the geochemical properties of deep groundwater sampled from the research area with granite, salt water intrusion, and the heavy weight of the canister. The results of safety analysis for the canister shows that the criticality for the normal case including uncertainty is the value of 0.816 which meets subcritical condition. Considering nation's 'Basic Plan for Electric Power Demand and Supply' and based on the scenario of disposing CANDU spent fuels in the first phase, the disposal system that the repository will be excavated in eight phases with the construction of the Underground Research Laboratory (URL) beginning in 2020 and commissioning in 2040 until the closure of the repository is proposed. Since there is close correlation between domestic HLW management plans and front-end/back-end fuel cycle plans causing such a great sensitivity of international environment factor, items related to assuring the non-proliferation and observing the international standard are showed to be the influential factor and acceptability of HLW management plans to be the influential factor of making decision for social acceptability of nuclear power

[en] To understand the long-term leach behavior of a borosilicate waste glass in a repository, the leaching experiment with three kinds of simulated borosilicate waste glasses has been carried out since the middle of 1997. The five years results indicate that a boron would be applied as an indicator of a long-term leachning of their borosilicate waste glasses and that their long-term leach rates have a tendency to be close to about 0.03g/m²-day even though their compositions and their ratios of the surface area to the volume of leachate are different

[en] The effect of exchangeable cation - Na⁺ and Ca²⁺ - on the diffusive transport of I^-, Sr²⁺ and ³H(as HTO) in compacted bentonite was examined using a through-diffusion method. Total intrinsic diffusion coefficients, D_i, were determined from the steady-state flux of the diffusants through the clays, and apparent diffusion coefficients, D_a, were obtained from the time lag technique. The clays were compacted to a dry bulk density of 1.3 Mg/m³, and Na-bentonite was saturated with a solution of 100 mol NaCI/m³ and Ca-bentonite with one of 50 mol CaCI₂/m³. The D_i values for all diffusants are 2 to 6 times higher in the Ca- than Na-clay. We attribute this to the larger quasicrystal, or particle, size of Ca- compared to Na-bentonite. Hence, Ca-bentonite has a greater proportion of relatively large pores; this was confirmed by Hg intrusion porosimetry. This means the diffusion pathways in Ca-bentonite are less tortuous than those in Na-bentonite. Moreover, in some cases the effective porosity, or the porosity available for diffusive transport, may be greater in Ca-bentonite. The Da values are inversely proportional to the distribution coefficients of the diffusants with the clays. (author)

[en] Rock properties as well as the influence of discontinies and DRZ, which should be considered for the analysis of structural stability of a high-level nuclear waste repository in deep deposit, were investigated. Also the mechanical criteria for assessing the results from computer simulations, which are carrying for the repository design, were defined

[en] The cold neutron produces by making thermal neutron passing through a volume of moderator installed into the reflector tank of the reactor. The HANARO CNS has adopted the liquid hydrogen as a moderator. The liquid hydrogen contained in the moderator cell evaporates due to the gamma heating. The hydrogen vapor goes up to the condenser, where it is re-liquefied then returns down to the moderator cell. This thermo-siphon loop can be only established under the very low temperature environment, which requires a method of a thermal insulation. The major performance of the vacuum system is to serve the normal operation vacuum level for thermal insulation of cryogenic components. The vacuum for the cryogenic insulation shall be at least lower than 10^-5 torr during the CNS normal operation. This report describes the results of detailed design of the vacuum system according to its design requirements. The summary derived from the detailed design are followings. - Non-safety class and Seismic II class (except for the equipment installed in the vacuum box) - Quality class is classified in T or S depending on the its design criteria - The vacuum system should be surrounded by the blanketing gas to isolate them from the air - The discharged gas from the vacuum system should be guided to the discharged gas collection tank and check whether the volume of hydrogen gas is less than 4% or not and then the discharged gas is released to the reactor hall. - In case, the hydrogen gas in the discharged gas collection tank is over than 4%, the outlet valve of the vacuum system should be automatically closed. - The pneumatic valves installed in the vacuum system should be actuated by the nitrogen gas and solenoid valves for operating the pneumatic valves should be placed to outside of the vacuum box. - It should be always required to measure the hydrogen leak from the vacuum box, which be filled with the blanketing gas, by using the proper hydrogen detector. - The design pressure of the vacuum box, the outer thing of double wall pipe, and the valve box is 3 MPa (a)

[en] For better cooperation in the field of radioactive waste management between Korea and Finland, the nuclear industry in Finalnd focusing its strategy and capabilities is studied. The strength of the Finnish program is in the site securing, disposal concept development with cost analysis, site investigation, and safety analysis. KAERI and Finnish delegates visited each other and held a joint seminar on the field of radioactive waste management. The future cooperation areas will be cost analysis and potential participation in the Onkalo project. Finnish DiP is effective to get consensus among many different stakeholders. Further independent study is recommended in this area