[en] KSC-4 transport cask and handling equipment should be safely stored and managed in the radiation controlled area. Therefore, it is necessary to approve the change of the PIEF’s warehouse to a radioactive waste storage facility. The purpose of this study is to obtain approval for change of license to the radioactive waste storage facility and to derive a plan for decontamination and decommissioning of KSC-4 cask. Basic data for decontamination and decommissioning of the transport cask were obtained by analyzing the decontamination and decommissioning technology for the spent fuel transport cask and nuclear facilities. The storage, decommissioning status and recyclability of unused cask were analyzed, and future treatment plan and recyclability of transport cask was suggested. The amount of radioactivity of the transport cask was predicted based on the results of measuring the internal radiation dose rate and contamination level of the KSC-4 cask. An analysis of the gas sampling method of the transport cask was performed and the concept of a gas sampling equipment was derived. An application for change of license has been submitted to the authority to convert PIEF’s warehouse into a radioactive waste storage facility. In addition, Q&A was conducted for two times of licensing review. After obtaining the change of license for the existing RG laboratory, the moving and installation of the equipment was completed, and applications for change of license the RI/NM laboratory were prepared. The results obtained from this study will be available as basic data for change of license to a radioactive idle equipment storage facility and decontamination & decommissioning of the KSC-4 cask

[en] KSC-4 transport cask and handling equipment should be safely stored and managed in the radiation controlled area. Therefore, it was necessary to approve the change of the PIEF’s warehouse to a radioactive waste storage facility. The purpose of this study is to secure basic data for change of license to a radioactive waste storage facility through the structural design, seismic analysis, fire hazard analysis and radiation environmental impact assessment for storage facility. Temporary storage facility was secured to safely store the KSC-4 transport cask. The storage facility was established as a temporary radiation controlled area, and transport cask and handling equipment were moved and installed. Structural design, seismic analysis, fire hazard analysis and radiation environmental impact assessment were performed for the radioactive waste storage facility. Structural safety of the storage facility was proved by the structural design and seismic analysis. Improvement item of fire protection equipment was derived by the fire hazard analysis. As a results of the radiation environmental impact assessment, the maximum personal exposure dose in the exclusion area boundary was evaluated within the allowable value. In order to secure a safe and legitimate storage facility of transport cask, an application for change of license has been submitted to the authority to convert PIEF’s warehouse into a radioactive waste storage facility

[en] This report presents the subchannel analysis in a dry cavity of PWR fuel shipping cask to verify the safety of dry type transport under normal transport condition. The cask can transport 4 PWR spent fuel assemblies with a burn-up of 38,000 MWD/MTU and a cooling time of 3 years. It is called KSC-4 shipping cask. Subchannel analysis was performed by using the COBRA-SFS computer code. The results of analysis were compared with the results of heat transfer test. Maximum fuel cladding temperature in the dry cavity of the cask was 238.deg. which is below the specified temperature of 532.deg. to maintain the integrity of the cask

[en] A fuel transport container for KiJang Research Reactor(KJRR) has been developed to transport fresh fuel assemblies and fission molly targets which are used for a research reactor built in Kijang. The KJRR fuel transport container is a type-A(F) container, which is defined in domestic and foreign regulations of a radioactive substance container. According to Nuclear Safety and Security Commission's notification, the container should meet the accident conditions defined in IAEA safety Standard Series, US NRC and etc. In this study, a structural integrity of the KJRR fuel transport container is evaluated by conducting computational analyses of 9-meter free drop and 1 meter puncture. It is confirmed that structural integrity of the KJRR fuel transport container can be maintained in the transportation accident condition. Hereafter, when the test model is produced, a safety test will be conducted and its result will be compared with the result of drop and puncture analyses.

[en] Highlights: • Thermal tests were performed to evaluate the heat removal performance of the concrete storage cask. • Passive heat removal system was well designed and worked adequately. • Half-blockage of the inlet has a relatively small effect. • Thermal integrity of the concrete is maintained under accident conditions. - Abstract: Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A concrete storage cask to safely store spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Moreover, the concrete storage cask must ensure that the temperatures of the spent nuclear fuel assemblies are maintained within the allowable values for normal, off-normal, and accident conditions. Therefore, the concrete storage cask must be designed to have heat removal capabilities with appropriate reliability. However, the thermal conductivity of concrete is not good and the allowable temperature of concrete is lower than that of steel. In this study, a thermal test was performed to evaluate the heat removal performance of the concrete storage cask under development by KORAD (Korea Radioactive Waste Agency), under normal and off-normal conditions. In addition, a thermal test was performed to evaluate the thermal integrity of the concrete under accident conditions. The heat transfer rate to the ambient atmosphere by convective airflow through the passive heat removal system of the concrete storage cask was found to reach 93.5% under normal conditions. Thus, it was confirmed that the passive heat removal system was well designed and worked adequately. In addition, the heat transfer rate to the ambient atmosphere by convective airflow through the passive heat removal system under off-normal conditions was estimated to reach 87.4%. Therefore, it was deduced that the half-blockage of the inlet openings has a relatively small effect on the maximum temperatures and temperature distributions. Moreover, no significant temperature differences were detected with respect to the location of the half-blockage of the inlet openings. This indicated that, the influence of the direction of the half-blockage of the inlet openings on the heat removal performance was minimal. Finally, the maximum temperature of the over-pack inner surface under accident conditions was measured as 103 °C, thus verifying that the thermal integrity of the concrete is adequately maintained under accident conditions

[en] Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the body of the test model through the heat transfer fin. The maximum temperatures of the neutron shielding at the part where the heat transfer fin was installed were 155 °C. However, those in the part where the heat transfer fin was not installed were 183 °C. The neutron shielding was therefore adequately protected by the heat transfer fin.

[en] An exponential experiment system composed of neutron detector, signal analysis system and neutron source (Cf-252, 4x10 7 n/s) has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron multiplication factors of PWR spent fuel assemblies. The neutron detector and source are inserted in the control rod guide tube of the C15 assembly in Kori unit 1 which was loaded in PIEF storage pool for the measurement of axial neutron flux distributions. The measurements are carried out when the detector or the neutron source is scanned in the axial direction and other one is fixed at 180 cm from the bottom end of the assembly. Both of the measured neutron distributions appeared in the similar exponential decay form and the exponential decay constants(γ) are determined to be 0.152 the for detector scanning and 1.65 for the source scanning, respectively. The neutron effective multiplication factor for the assembly is estimated to be 0.480 and 0.441 for both exponential decay constants, respectively

[en] In this study, a thermal analysis of the cask with bird screen meshes has been performed using a porous media model. The overpack consists of a structural material, a concrete shielding, and a ventilation system. Heat is removed from the cask to the environment by a passive means only. Air inlet and outlet ducts are installed at the bottom and top of the cask for a ventilation system. Bird screen meshes are installed at the air inlet and outlet ducts to inhibit intrusion of debris from the external environment. The presence of this screens introduce an additional resistance to air flow through the ducts. Five types of meshes for bird screen were considered in this study. The bird screen meshes at the inlet and outlet vents reduce the open area for flow by about 44 - 79 %. Flow resistance coefficients for porous media model were deduced from the fluid flow analysis of bird screen meshes. Thermal analyses for the concrete cask have been carried out using a porous media model. The analysis results agreed well with the test results. Therefore, it was shown that the porous media model for the screen mesh was established to estimate the cask temperatures

[en] Sealed sources have to conducted the tests be done according to the classification requirements for their typical usages in accordance with the relevant domestic notice standard and ISO 2919. After each test, the source shall be examined visually for loss of integrity and pass an appropriate leakage test. Tests to class a sealed source are temperature, external pressure, impact, vibration and puncture test. The environmental test conditions for tests with class numbers are arranged in increasing order of severity. In this study, the apparatus of tests, except the vibration test, were developed and applied to three kinds of sealed source. The conditions of the tests to class a sealed source were stated and the difference between the domestic notice standard and ISO 2919 were considered. And apparatus of the tests were made. Using developed apparatus we conducted the test for ¹⁹²Ir brachytherapy sealed source and two kinds of sealed source for industrial radiography. ¹⁹²Ir brachytherapy sealed source is classified by temperature class 5, external pressure class 3, impact class 2 and vibration and puncture class 1. Two kinds of sealed source for industrial radiography are classified by temperature class 4, external pressure class 2, impact and puncture class 5 and vibration class 1. After the tests, Liquid nitrogen bubble test and vacuum bubble test were done to evaluate the safety of the sealed sources

[en] This paper describes the beneficial effect of weldment imperfection of the cask impact limiter, by applying intermittent-weld, for impact energy absorbing behavior. From the point of view of energy absorbing efficiency of an energy absorber, it is desirable to reduce the crush load resistance and increase the deformation of the energy absorber within certain limit. This paper presents the test results of intermittent-weldment and the analysis results of cask impacts and the discussions of the improvement of impact mitigating effect by the imperfect-weldment. The rupture of imperfect weldment of an impact limiter improves the energy-absorbing efficiency by reducing the crush load amplitude without loss of total energy absorption. The beneficial effect of weldment imperfection should be considered to the cask impact limiter design. (author)