In this study, we focus on the feasibility of structural topology optimization for a steel-timber composite beam design of optimally allocating glue-laminated timbers into a web with openings under the condition of given steel flanges. The motivation of this study is to topologically take maximal stiffness harmonizing both tension and compression performance of the steel-timber composite beam and become the eco-frandly timber design for buidling members. As a result of this study, the key web-openings allocation becomes triangle spaces, i.e., empty or no materials, of optimal topologies of both a pure timber plate and a steel flange-web timber plate without web-openings. Several applicable examples verify the effectiveness of topology optimization for steel-timber beams with web-openings.

Recently, machine learning is widely used to solve optimization problems in various engineering fields. In this study, machine learning is applied to development of a control algorithm for a smart control device for reduction of seismic responses. For this purpose, Deep Q-network (DQN) out of reinforcement learning algorithms was employed to develop control algorithm. A single degree of freedom (SDOF) structure with a smart tuned mass damper (TMD) was used as an example structure. A smart TMD system was composed of MR (magnetorheological) damper instead of passive damper. Reward design of reinforcement learning mainly affects the control performance of the smart TMD. Various hyperparameters were investigated to optimize the control performance of DQN-based control algorithm. Usually, decrease of the time step for numerical simulation is desirable to increase the accuracy of simulation results. However, the numerical simulation results presented that decrease of the time step for reward calculation might decrease the control performance of DQN-based control algorithm. Therefore, a proper time step for reward calculation should be selected in a DQN training process.

PURPOSES : This study evaluates the long-term performance of the asphalt overlay designed by the Seoul pavement design method which determines overlay thickness by considering existing pavement conditions, traffic volume, and bearing capacity of the pavement. METHODS : A total of 76 sections including 17 control sections and 59 design sections were constructed under various traffic conditions, overlay thicknesses and asphalt mixtures. The performance of the pavements has been monitored up to 60 months in terms of surface distresses, rutting, and longitudinal roughness. The service life of the pavements was estimated to be the period when the Seoul pavement condition index (SPI) becomes 6.0, i.e., a rehabilitation level. RESULTS : Overall, the service life of the pavements was 72 months in the control and 120 months for the design sections. For relatively thinner overlay sections than designed, the service life reduced significantly; 36 months for 15cm thick overlay and 120 months for 25cm thick overlay. The service life of the pavement in the bus-only lane was 78 months, which is 30 months shorter than that in mixed-traffic lanes. Out of the bus-only lanes, 56% of the pavement along bus stop was deteriorated early to be a poor condition while only 2% of the pavement in a driving lane was degraded to be poor. The overlay with Stone Mastic Asphalt (SMA) in the wearing surface had 38% longer life than that with conventional dense graded mixtures. CONCLUSIONS : Most of the overlays sections designed by the Seoul pavement design method were expected to survive 10 years, except for bus-only lanes. The control sections having 5 to 10 cm thick overlays showed significant lower performance than the design sections. Thus proper thickness and materials considering the characteristics of existing pavement and traffic volumes should be applied to secure the service life of overlays.

Designing and producing a low-cost, high-current-density electrode with good electrocatalytic activity for the oxygen evolution reaction (OER) is still a major challenge for the industrial hydrogen energy economy. In this study, nanostructured Fe-doped CuCo(OH)2 was discovered to be a precedent electrocatalyst for OER with low overpotential, low Tafel slope, good durability, and high electrochemically active surface sites at reduced mass loadings. Fe-doped CuCo(OH)2 nanosheets are made using a hydrothermal synthesis process. These nanosheets are clumped together to form a highly open hierarchical structure. When used as an electrocatalyst, the Fe-doped CuCo(OH)2 nanosheets required an overpotential of 260 mV to reach a current density of 50 mA cm−2. Also, it showed a small Tafel slope of 72.9 mV dec−1, and superior stability while catalyzing the generation of O2 continuously for 20 hours. The Fe-doped CuCo(OH)2 was found to have a large number of active sites which provide hierarchical and stable transfer routes for both electrolyte ions and electrons, resulting in exceptional OER performance.

Due to the Fukushima nuclear accident, a large amount of radioactive material was released into the atmosphere, and consequently, it spread over a wide area and was deposited into the soil. As a result of this, a wide area of radioactive contamination site was created. Due to the contaminated site, the need for research on various exploration platforms for efficient situation management and field response is being emphasized. Backpack-type radiation survey & monitoring equipment is useful for creating a contamination maps containing information such as Dose Rate, Radionuclide, Activity Concentration accompanied by spatial analysis when performing a Scan Survey that moves with a backpack on a wide area site. contamination maps are based on accurate radiological characteristic information. However, there is a problem in that the accuracy of the evaluation results is lowered due to changes in environment conditions or the variability of the dose rate and counting rate during scan survey. This problem should be solved by applying the influence of each variable to the underlying data. However, prior to this, it is most important to prepare the base underlying first. And this can be obtained through evaluation of detection performance through static survey. Therefore, in this study, the change in detection efficiency for the measurement height and radius of the backpack-type radiation survey & monitoring equipment based on the 3"×3" NaI(Ti) detector was evaluated. First, the height of the source and Backpack-type radiation survey & monitoring equipment was evaluated from 0 cm to 1 m, which is the height of the soil and detector when an adult male wears a backpack. The experiment was conducted using the 137Cs (383 kBq) point source, which is a nuclide mainly detected at the contaminated site. The measurement time was measured five times per one minute, considering that it was backpack-type equipment and a future scan survey. In addition, in order to evaluate the detection radius, the measurement was performed by changing the measurement distance up to 5 m at intervals of 50 cm. As a result of evaluating the detection performance of the backpack-type radiation surveys & monitoring equipment, it was confirmed that increasing the detection height and radius reduces the count rate in the form of an exponential function. In addition, it was confirmed that the detected radius varies depending on the height. Based on these results, we plan to conduct additional research to understand the scan survey and its sensitivity to various factors. Through this, the company plans to develop various models for exploring the site by improving the accuracy of backpack-type radiation surveys & monitoring equipment.

The decommissioning of Kori unit 1 is just around the corner. Accordingly, it is required to construct a hot cell facility for decommissioning nuclear power plants to analyze the characteristics of intermediate-level waste and low-level waste generated in the decommissioning process. In this study, a Design Base Accident (DBA) scenario of the facility is developed. To identify and characterize potential hazards at the facility, a Preliminary Fact Sheet (PFS) is filled out and consider external events in consideration of the surrounding site environment. The external event screening and evaluation method is based on the external event evaluation method covered in the probabilistic risk assessment. In PFS, only natural and artificial hazards that may have a meaningful impact on the facility are considered as the sources of the accident, and accident prevention and mitigation systems, etc., which exist in each compartment or facility, are described. Based on PFS and external events, potential hazard assessment is systematically performed using each potential hazards, impact and defense function identified using the preliminary hazard analysis (PHA) methodology. The potential hazard analysis methodology applied to this assessment is a qualitative assessment method consistent with the US DOE Hazard Analysis methodology (DOE 1992b; DOE 1994b). After that, the potential mitigation functions that can be used under normal, abnormal and accident conditions are examined, and the contribution of public and workers to safety is evaluated. The results of the PHA are basic data that prioritize potential hazards and can be used to develop potential accident scenarios. Among potential hazards generally considered for non-reactor facilities, only possible accidents during operation of the facilities are selected as potential hazards. The level of potential hazards is obtained by qualitatively examining the frequency and consequence estimates for each hazard or accident scenario developed in PHA. Based on the results of the potential hazards assessment, representative accidents that require further quantitative analysis are screened. Selected accidents are DBA and are the most dangerous and most significant impacts on workers.

In domestic nuclear power plants, drums of concentrated radioactive waste solidified with paraffin that do not meet radioactive waste disposal standards are stored temporarily. In this paper, the design of a machine that separates these paraffin drums into paraffin and concentrated waste using heating vaporization and pressure difference is described. The separation process is as follows. First, the paraffin solid is indirectly heated by heating the outside of the drum. The paraffin solid is partially melted to increase the fluidity and is easily detached from the drum. The detached solid is transferred to the melting tank, and further heated in the melting tank. When the temperature is sufficiently high, paraffin is melted and becomes a mixture of liquid paraffin and concentrated waste homogeneously. The mixed solution is transferred to a paraffin recovery vessel and further heated. The vaporization point of paraffin is 370°C under atmospheric pressure, and becomes lower depending on the pressure decreasing in the vessel. The vaporization point of the paraffin is a relatively low value compared to the radioactive elements in the concentrated waste, and therefore only paraffin would be vaporized. A paraffin transfer pipe is installed on the upper part of the paraffin recovery vessel, and is connected to another tank called the paraffin capture vessel. The pressure of the paraffin capture vessel is reduced (i.e. vacuum condition), only gaseous paraffin is transferred to the paraffin capture vessel by the pressure difference. When the paraffin capture vessel is cooled below the vaporization point of the paraffin, the paraffin is liquefied or solidified, and only the paraffin is recovered. Based on the above process, the solidified paraffin could be separated into pure paraffin and concentrated waste. However, if a radioactive element with a lower vaporization point than paraffin exists in the concentrated waste, it may be mixed with paraffin and separated together. Therefore, it is necessary to measure the radioactivity or radiation dose rate for the separated paraffin, and to verify that it is sufficiently low. If necessary, additional separation process may be considered for removing radioisotopes from the paraffin.

In this work, we introduce a 100 kW class mobile plasma melting system designed for non-combustible radioactive wastes treatment. To ensure mobility, the designed system consists of two 24-ft commercial containers, each in charge of the plasma utilities and melting process. In the container for plasma utilities, a 100 kW class DC power supply is installed together with a chiller and gas supply system whereas the container for melting process has a transferred type arc melter as well as off-gas treatment system consisting of a heat exchanger, filtrations, scrubber and NOx removal system. As a heat source for a transferred type arc melter, we adopted a hollow electrode plasma torch with reverse polarity discharge structure. Detailed design for a 100 kW class mobile plasma melting system will be presented together with the main specifications of the components. In addition, the basic performance data of the melting system is also presented and discussed.

Since July 2021, the Korea Radioactive Waste Agency has been conducting a safety case development study for the Korean deep geological repository program. The safety case includes generating scenarios in which radioactive materials from spent nuclear fuel repository reach the human biosphere by combining selective FEPs (Features, Events, and Processes). This safety case should be able to transparently explain the process in which conclusions have been drawn not only to stakeholders but also to the public by presenting safety arguments. The scenario development stage consisting of FEP screening, scenario generation, and uncertainty analysis procedures should have a database management system. Database management system was performed in countries such as Sweden, which obtained approval for the construction of spent nuclear fuel repositories, and the United States, where various preliminary research was carried out. Korea Atomic Energy Research Institute also has experience in designing and operating its own database, which has conducted preliminary research on disposal of the spent nuclear fuel. Currently, the safety assessment of the Korean spent nuclear fuel repository is in the early stages of research, but it is necessary to set up a basic framework for database design while the collection of FEP data from domestic and international preliminary studies is under development, and it is advantageous for efficient database construction and operation. Therefore, this paper presents the current status of database design considering completeness and transparency from the FEP screening stage to the scenario development stage in the safety assessment process of the Korean spent nuclear fuel repository. In this process, the functional requirements that the database should provide, the database schema capable of implementing them, and simple examples are presented together. The objectives of this database design are flexible FEPs management, high integrity and consistency, and expandability for linking with the safety case database. The FEP data to be inputted into the database includes a list of major opened FEPs, including International FEPs from Nuclear Energy Agency, which were referred for PFEPs (Project-specific FEPs), and PFEPs applied to POSIVA's Olkiluoto repository. As an additional function, queries from the database are used to visually express the process of deriving scenarios through Rock Engineering System, a widely known scenario generation methodology.

This paper intends to present considerations on the question of what is the “load standard” or “design load” for integrity evaluation under normal transportation conditions and what type of design load is good for users. This suggests a direction for subsequent research on producing design loads that transport business companies can utilize without difficulty. Several studies have been conducted to evaluate the integrity of spent nuclear fuel during normal transportation. A representative study recently conducted is the Multi-modal Transportation Test (MMTT) conducted using a commercial spent nuclear fuel cask by US DOE in 2017. In Korea, additional transport tests were planned to acquire sufficient test data under the conditions of road and sea transport considering the Korean situation. As a result, road transport tests were carried out in 2020 and sea transport tests were carried out in 2021. In the road transport test, a driving test that simulates various road conditions and a test that cycled a 4.5 km road eight times were performed. In most cases, the maximum acceleration of less than 1 g occurred, and the maximum strain was less than 48 με. For the sea transport test, the magnitude of both the maximum acceleration and the maximum strain were lower than those in the road transport test. We concluded tentatively that the integrity of spent fuel under normal conditions of transport was satisfactory with a large margin. However, when the storage business is realized and the transport of spent fuel becomes visible, the storage and transport business companies will have to prove the maintenance of the integrity of the spent fuel under normal transport conditions at the request of the regulatory agency. The transport business companies can transport the spent nuclear fuel by using different types of transport casks and different types of trucks and ships from those used in the tests mentioned above. However, it is absurd to have to prove the integrity of spent nuclear fuel by performing expensive tests again. Therefore, in this study, the design load that can be used by transport business companies is to be presented. The design load to be presented should satisfy the following requirements. The design load should be applicable including some differences in the transport cask or transport system, or different design loads should be presented according to the differences. The location where this design load is applied is to be specified (e.g. fuel rod, basket, internal structure). Requirements according to the operating speed of the transport system should be presented together. The type of design load is to be presented (e.g. PSD, SRS, FDS etc.). Other types of standards may be presented. For example, a speed limit for a vehicle carrying spent nuclear fuel may be suggested, or a speed limit for a vehicle passing through a speed bump may be suggested. In order to present such a reliable design load, a multi-axis vibration excitation shaker table test will be carried out. Though this shaker table test, the behavior of the nuclear fuel assembly is closely evaluated by applying the data obtained from the road and sea transport tests previously performed as an input load. In addition, FDS (Fatigue Damage Spectrum) will be produced and applied to experimentally evaluate the durability of fuel assemblies under normal transport conditions.