In accordance with the notification of the Nuclear Safety and Security Commission (NSSC), environmental impact assessments around nuclear power plants are conducted annually and the results are disclosed to the public. KHNP evaluates the dose of residents around nuclear power plants using the K-DOSE60 program that reflects ICRP-60. K-DOSE60 calculates the expected exposure dose for residents by modifying the atmospheric dispersion and deposition factors evaluation module (XOQDOQ), gaseous effluent evaluation module (GASDOS) and liquid effluent evaluation module (LIQDOS) developed by the US NRC. The current evaluation program is the Bounding Assessments method, which evaluates under the assumption that residents reside at the exclusion area boundary (EAB), and has a disadvantage in that the estimated exposure dose is evaluated too conservatively. In the EPRI, instead of the conservative method that is conventionally performed for the residents’ dose evaluation method, a plan to improve the accuracy of the dose evaluation reflecting the site characteristics was reviewed. In addition, improvements were derived through the review of NPPs operation status, experience cases and the latest technology.

The Co-60 is a radioactive material widely used in domestic and foreign medical, industrial, health and research fields. Currently, world market for the Co-60 is about 80 MCi/yr and is expected to grow to 150 MCi/yr by 2025. For the Co-60, Nordion of Canada occupies about 80% of the world market. In the case of Korea, a small amount of sources with relatively low radioactivity intensity are produced using research reactors, but most of the Co-60 is entirely dependent on imports. Accordingly, although the technical feasibility of the Co-60 production technology using the PHWR was evaluated, it was evaluated as a negative result on the additional construction of a hot cell, core management, safety analysis and economic feasibility. Canada, the main producer of the Co-60, is also conducting research on the Co-60 production technology using PWR with GE-Hitachi and Westinghouse as the number of PHWR is expected to decrease. In Korea, it is necessary to preoccupy the Co-60 production technology and auxiliary technology using the PWR by utilizing excellent technology, and active research is being conducted to secure unique nuclear power technology that does not depend on foreign countries. Therefore, in this study, the thickness and weight of the radioactive shielding required for handling (transport) of Co-60 produced using the PWR were calculated.

The off-site dose calculation is regularly carried out at the nuclear power plants in order to evaluate off-site dose from gaseous and liquid effluent during normal operation. In 2009, the off-site calculation program (K-DOSE60) was developed in accordance with ICRP-60 by KHNP. This software needs meteorological data, gaseous and liquid effluent data, and various other input parameters to evaluate off-site dose. As a result, it takes a certain amount of time for the user to enter accurate input data and verify calculated results, and it is difficult to intuitively determine them because of providing textbased calculated results. Therefore, in this study, the improvement of the calculation program was considered so that a more reliable and effective evaluation could be performed when calculating the off-site dose. The main improvements of the off-site dose calculation program (ODCP) are as follows. First, it is developed as the network-based program to link with meteorological data, and gaseous and liquid effluent data to remove input errors and simplify data transfer. Second, through validation process of input data, input errors are eliminated. Third, the input data and calculated results are visually provided so that the user can easily determine the evaluation results. Fourth, database of input and calculated results is constructed to facilitate evaluation result history management.

The buffer material plays a role in preventing the excessive rise in temperature generated from the high-level radioactive waste by dissipating the decay heat to the rock. For this reason, the buffer material must have thermal properties to ensure the performance of the deep geological repository. This study measured the thermal conductivity of sand-bentonite according to the mixing ratio to improve the thermal properties. The compacted buffer was manufactured with a sand-bentonite mixing ratio of 6:4, 7:3, and 8:2 with 9 to 12% water content. As a result, the thermal conductivity increases as the ratio of sand increases. As a further study, it is necessary to experiment on whether sand-bentonite’s hydraulic, mechanical, and chemical performance is suitable for the stable operation of a repository.

A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. Among these, the bentonite buffer is one of the most important components to assure the safe disposal of high-level radioactive waste (HLW). As the bentonite buffer is installed as a block type, it is important to fabricate homogeneously. Generally, floating die method and cold isostatic press (CIP) method are used to fabricate bentonite blocks. In this paper, two bentonite blocks were produced using float die method at first, and CIP method was additionally applied to just one block. After that, several samples were cored from two blocks. The dry density and water content of several samples produced from two blocks were measured.

PWR spent nuclear fuel generally showed an oxide film thickness of 100 um or more with a combustion rate of 45 MWD/MTU or higher, while CANDU spent nuclear fuel with an average combustion rate of about 7.8 MWD/MTU had few issues related to hydride corrosion. Even based on the actual power plant data, it is known that the thickness of the oxide film is 10 μm or less on the surface of the coating tube, and brittleness caused by hydride is shown from the thickness of the oxide film of about 80 μm, so it is not worth considering. However, since corrosion may be accelerated by lithium ions, lithium ions may be said to be a very important factor in controlling the hydro-chemical environment of heavy water. Lithium has a negative effect on the corrosion of zirconium alloys. However, since local below 5 ppb to prevent corrosion. maintained at a concentration between 0.35 and 0.55 ppm. Hydrogen is known to have a positive effect by suppressing radioactive decomposition of the coolant and suppressing cracks in nickelbased alloys. However, too much hydrogen can produce hydride in a pressure tube composed of Zr-2.5Nb, so DH (Disolved Hydrogen) maintains the range of 0.27–0.90 ppm. pH and conductivity are completely determined by lithium ions, and DH can be completely removed below 5 ppb to prevent corrosion. Therefore, for cladding corrosion simulation of the CANDU spent nuclear fuel, a hydrochemical of the equipment, not 310°C, and 14 uS·Cm−1 is targeted as conditions for corrosion acceleration. In addition, for acceleration, the temperature was set to 345°C (margin 10°C), which is the maximum accommodation range of the equipment, not 310°C.

The manufactured nuclear fuel assembly is loaded into the nuclear reactor after the core design, and is finally discharged to the wet storage pool after depletion for 3 cycles. The discharged spent nuclear fuel is transported and stored in a dry storage system at the on-site of the nuclear power plant, which is cooled by natural convection, and undergoes final disposal or reprocessing through an intermediate dry storage facility. In this series of processes, the characteristics of the final product, the spent fuel, vary depending on the environmental conditions, so it is essential to manage each history data to verify the long-term integrity of the spent nuclear fuel. In this paper, safety information on spent nuclear fuel is described in order to establish technical requirements that should be considered in each stage of storage, transport, reprocessing, and disposal of spent nuclear fuel. Comprehensive safety information on spent nuclear fuel is basically calculated from basic information that considers characteristic information that can be obtained through the manufacture and design of nuclear fuel assemblies, operation history in a nuclear reactor, and location history in a wet storage pool. It can be divided into secondary production information (SF Burnup, Nuclide Inventory, etc.) and tertiary integrity-related information obtained through cladding inspection during spent fuel storage. KHNP produces this multi-layered information according to the production stage and manages it through the comprehensive management system of the spent nuclear fuel, and safety information with some errors is not only improved through re-verification but also continuously updated. In this paper, the spent nuclear fuel safety information was derived based on various information calculated in the entire process of being discharged and managed in a wet storage pool, including new fuel manufacturing information and depletion history. Such safety information will be used as basic data for long-term safe management of spent nuclear fuel, and will be continuously produced and managed. In the future, additional discussions will be held on the safety information of the spent nuclear fuel through consultation with KORAD and regulatory agencies.

수처리 및 의약바이오 분야에서 유효물질 분리에 활용되고 있는 알루미나 중공사 분리막은 얇은 두께로 인해 취 급 및 적용시 쉽게 파괴되는 단점이 있기 때문에 분리막의 강도를 100 MPa 이상으로 향상시키기 위한 연구가 필요하다. 본 연구에서는 나노입자의 함량을 0, 1, 3, 5 wt%로 증가시켰을 때 제조된 중공사 분리막의 특성을 평가하였다. 그 결과, 나노입 자의 함량이 증가함에 따라 중공사 분리막의 강도는 79 MPa에서 115 MPa로 증가하였으며, 밀도는 1.76 g/m3에서 1.88 g/m3 으로 증가하였고 기공률과 평균기공크기는 각각 51%에서 48%로, 416 nm에서 352 nm로 감소한 것을 확인하였다. 스폰지구 조가 발달하고 스폰지구조의 기공크기가 향상된 알루미나 중공사 분리막은 100 MPa 이상으로 기계적 강도가 향상되었으며, 약 100000 GPU의 높은 질소 투과도 및 약 3000 L/m2h의 높은 물 투과도를 나타내었다. 따라서, γ-알루미나 나노입자를 소 결조제로 첨가하는 것은 α-알루미나 중공사 분리막의 기계적 강도를 효과적으로 증진시키고 높은 투과성능을 유지할 수 있 는 매우 유효한 방법임을 확인하였다.