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.

The purpose of this study is to evaluate dogs' sociality toward human strangers in the absence of an owner by analyzing changes in dogs' behavior during a task of making eye contact with an experimenter to obtain snacks. A total of 17 dogs were divided into groups of high sociality (HS; n = 10, 4.4 ± 3.87 years) and low-sociality (LS; n = 7, 3.71 ± 2.06 years). A comparison of the average frequency of five behavioral types-fear-appeasement behaviors (P<0.001), sociability-related behaviors (P<0.001), stress-related behaviors (P<0.05), destruction (P < 0.001), and vocalization (P < 0.001)-between the groups showed a significant difference in all five categories. Together, these results suggest that dogs with high sociality are less exposed to various stresses and have a higher ability to adapt to new environments than dogs with low sociality. This can predict dogs' adaptability to a new environment and positive outcomes in their daily life with the owner.

This study aimed to evaluate the effects of four types of environmental enrichment on the improvement of companion dogs' behavioral problems due to separation anxiety. A total of 21 dogs of various breeds were included in the study. Data were collected to investigate the behaviors associated with anxiety in dogs, including vocalization, elimination, escape attempts, and destructiveness. A first stage, in which the dog and owner were together (P0), lasted 15 min, and a second stage, in which the dog and owner were separated (P1), lasted 15 min. After the dog and owner were separated (P1), the third stage (P2), during which the environment was enriched, lasted 20 min, and the fourth stage, following environment enrichment (P3), lasted 15 min. The results of the study indicated that compared to P0, the frequency of problematic behavior was highest during the 15 min following separation from the owner (P1). Following environmental enrichment, the average frequency of problematic behaviors in P2 decreased (P < 0.001) compared to P1. Environmental enrichment can also be used appropriately in the case of companion dogs, including shelter dogs or experimental dogs that use a limited kennel, and is a particularly effective means of improving the quality of life of dogs.

Graphite Isotope Ratio Method (GIRM) can be used to estimate plutonium production in a graphite-moderated reactor. This study presents verification results for the GIRM combined with a 3-D polynomial regression function to estimate cumulative plutonium production in a graphite-moderated reactor. Using the 3-D Monte-Carlo method, verification was done by comparing the cumulative plutonium production with the GIRM. The GIRM can estimate plutonium production for specific sampling points using a function that is based on an isotope ratio of impurity elements. In this study, the 10B/11B isotope ratio was chosen and calculated for sampling points. Then, 3-D polynomial regression was used to derive a function that represents a whole core cumulative plutonium production map. To verify the accuracy of the GIRM with polynomial regression, the reference value of plutonium production was calculated using a Monte-Carlo code, MCS, up to 4250 days of depletion. Moreover, the amount of plutonium produced in certain axial layers and fuel pins at 1250, 2250, and 3250 days of depletion was obtained and used for additional verification. As a result, the difference in the total cumulative plutonium production based on the MCS and GIRM results was found below 3.1% with regard to the root mean square (RMS) error.

While Shandong Province is the largest historical and cultural and the first agricultural province in China, in recent years, due to the accelerated urbanization and modernization process, the conflict between traditional culture and modern civilization has had a serious impact on rural culture, and the ancient architecture among the villages in Shandong has also caused certain damage. These ancient buildings scattered in various villages in Shandong are important historical and cultural resources, the embodiment of Shandong traditional culture, and have an important role in the revitalization of rural culture and the enhancement of rural cohesion. However, research on the subsequent survey, conservation and restoration, and tourism development of ancient buildings in Shandong has lagged behind, and the role of these buildings in the inheritance of traditional Shandong traditional culture has not been given due attention, and their role in the construction of rural culture in the new era of Shandong has been seriously neglected. Tourism utilization is generally considered to be the preferred way to solve the conservation problems of ancient buildings. This study argues that the cultural revitalization of rural culture can be driven by the tourism and cultural industry of ancient buildings, and it can also provide a strong and sustainable power source for the rural revitalization strategy of the whole Shandong Province.

The dependencies of the chemical element abundances in stellar atmospheres with respect to solar abundances on the second ionization potentials of the same elements were investigated using the published stellar abundance patterns for 1,149 G and K giants in the Local Region of the Galaxy. The correlations between the relative abundances of chemical elements and their second ionization potentials were calculated for groups of stars with effective temperatures between 3,764 and 7,725 K. Correlations were identified for chemical elements with second ionization potentials of 12.5 eV to 20 eV and for elements with second ionization potentials higher than 20 eV. For the first group of elements, the correlation coefficients were positive for stars with effective temperatures lower than 5,300 K and negative for stars with effective temperatures from 5,300 K to 7,725 K. The results of this study and the comparison with earlier results for hotter stars confirm the variations in these correlations with the effective temperature. A possible explanation for the observed effects is the accretion of hydrogen and helium atoms from the interstellar medium.

In this work, the problem of resonance caused by some gravitational potentials due to Mercury and a third body, namely the Sun, together with some non-gravitational perturbations, specifically coronal mass ejections and solar wind in addition to radiation pressure, are investigated. Some simplifying assumptions without loss of accuracy are employed. The considered force model is constructed. Then the Delaunay canonical set is introduced. The Hamiltonian of the problem is obtained then it is expressed in terms of the Deluanay canonical set. The Hamiltonian is re-ordered to adopt it to the perturbation technique used to solve the problem. The Lie transform method is surveyed. The Hamiltonian is doubly averaged. The resonance capture is investigated. Finally, some numerical simulations are illustrated and are analyzed. Many resonant inclinations are revealed.