[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] 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] The purpose of this paper is to identify the type of structural load of the upper mechanism and to establish the design criteria of normal condition for high capacity CANDU spent fuel cask. And in the conceptual design review stage, computational modeling was performed for structural analysis according to the normal condition load. In this study, it is necessary to evaluate the structural safety of the upper mechanism because the flask stacking and the normal dropping condition that can occur in the normal operation of the spent fuel large capacity transport cask for heavy water reactor should have the safe function of upper mechanism. Dynamic analysis for 0.3 m normal condition drop had also the final residual deformation under the housing design criteria gap but at impact moment, maximum deformation is close to the design criteria gap and the stress of housing are beyond the elastic limit. Therefore, a part of the housing structures shall be reinforced not to exceed its elastic range.

[en] Transport package for radioactive material should be structurally safe under puncture drop condition and its safety should be verified by test and numerical analysis. Most finite element analyses for puncture drop have been performed without modeling the impact limiter since failure is occurred in the materials of the impact limiter. This paper presents a new modeling methodology, where an element is eroded in case that the material's failure criteria are reached at the element's integration point, to investigate the effect of the impact limiter in the puncture process. The effectiveness of the proposed scheme is shown through the puncture drop analysis of hotcell transport cask, which is under design in KAERI. The results show that about 80 percent of the total impact energy is absorbed due to the deformation of impact limiter. Using the present method the puncture drop can be analyzed more accurately, but it would give conservative results compared to the actual test condition.

[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] In robust design, the mean and variance of design performance are frequently used to measure the design performance and its robustness under uncertainties. In this paper, we present the Gauss-type quadrature formula as a rigorous method for mean and variance estimation involving arbitrary input distributions and further extend its use to robust design optimization. One dimensional Gauss-type quadrature formula are constructed from the input probability distributions and utilized in the construction of multidimensional quadrature formula such as the tensor product quadrature (TPQ) formula and the univariate dimension reduction (UDR) method. To improve the efficiency of using it for robust design optimization, a semi-analytic design sensitivity analysis with respect to the statistical moments is proposed. The proposed approach is applied to a simple bench mark problems and robust topology optimization of structures considering various types of uncertainty

[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] 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] Burnups for 36 points of five rods in the G23 assembly of Kori unit 1 have been determined on the basis of gamma-ray spectrometric measurement of two isotopic ratios, Cs-134/Cs-137 and Eu-154/Cs-137 in combination with the results calculated by the SCALE4.4 SAS2H module. Benchmarking of the SAS2H module has been done for the existing experimental data of Cs-13134, Cs-137 and Eu-154 isotopic compositions in PWR spent fuel. The gamma ray counts of two isotopic ratios have been corrected with their branching ratios, decay rates and energy dependent counting efficiencies in order to get true ratios. The energy dependent counting efficiencies have been determined as a quadratic equation based on the gamma ray counts for Cs-134 and Eu-154 at fourth energy points. Finally, burnups have been determined by putting true ratios of two isotopic ratios to their burnup-to-ratio fitting functions, respectively. Then the measured burnups have been compared with the declared burnup by the nuclear power plant. It is revealed that burnups determined from Cs-134/Cs-137 are agreeable with the declared burnups in most cases within about 12% error except a measuring point of C13, one of G23 fuel rods. In the case of Eu-154/Cs-137, the measured burnup is much lower than the declared burnup, which seems to be derived from system errors. (author)

[en] The concrete overpack of the cask provides radiation shielding as well as physical protection for inner canister against external mechanical shock. When the overpack undergoes a severe missile impact which might be caused by tornado or aircraft crash, it should sustain minimal level of structural integrity so that the radiation shielding and the retrievability of canister are maintained. Empirical formulas have been developed for the evaluation of concrete damage but those formulas can be used only for local damage evaluation and not for global damage evaluation. In this research, a series of numerical simulations and tests have been performed to evaluate the damage of two types of concrete overpack segment models under high speed missile impact. It is shown that appropriate modeling of material failure is crucial in this kind of analyses and finding the correct failure parameters may not be straightforward. When comparing the simulation results with the test results, it is shown that neither setting, case 1 and 2 provides results with consistent agreement with test results. That is, case 1 setting is more close to reality in Type 1 model analysis, but for Type 2, case 2 setting provides more close results to the reality. In both the case, not enough deformation is predicted by simulation compared to the tests. Weak failure and eroding criteria give larger penetration depth with insufficient overall damage due to energy loss with element erosion