• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1511-1518
• /
• 2021

In most countries, the thermal criteria for the engineered barrier system (EBS) is set to below 100 ℃ due to the possible illitization in the buffer, which will likely be detrimental to the performance and safety of the repository. On the other hand, if the thermal criteria for the EBS increases, the disposal density and the cost-effectiveness for the high-level radioactive wastes will dramatically increase. Thus, fundamentals on the thermal behavior of the buffer under the elevated temperatures is of crucial importance. Yet, the behaviors at the elevated temperatures of the bentonite under groundwater-saturated conditions have not been reported to-date. Here, we have developed an in-situ synchrotron-based method for the thermal behavior study of the buffer under the elevated temperatures (25-250 ℃), investigated dspacings of the montmorillonite in the Korean bentonite (i.e., Ca-type) at dry and KURT (KAERI Underground Research Tunnel) groundwater-saturated conditions (KJ-ii-dry and KJ-ii-wet), and compared the behaviors with that of MX-80 (i.e., Na-type, MX-80-wet). The hydration states analyzed show tri-, bi-, and mono-hydrated at 25, 120, and 250 ℃, respectively for KJ-ii-wet, whereas tri-, mono-, and de-hydrated at 25, 150, and 250 ℃, respectively for MX-80-wet. The Korean bentonite starts losing the interlayered water at lower temperatures; however, holds them better at higher temperatures as compared with MX-80.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1479-1482
• /
• 2021

Research reactors are typically well-suited for outreach activities at different levels. However, unplanned seeking to increase the utilization of a research reactor may result in weakening the nuclear security of this facility. Research reactor staff might be in shortage of a functional nuclear security culture; specifically, there might be a conviction that the necessities of research can be given the priority over consistence with security procedural requirements. Research reactors are usually parts of bigger institutes or research labs of different activities. Moreover, the employments of research reactors are usually with the purpose that easy entry to the reactor premises is fundamental. So, they could be co-situated in places with different sorts of activities, mostly under similar security arrangements. The co-area of research reactor offices among different kinds of research labs introduces explicit security issues, the effects of which should be viewed as when building up a nuclear security framework. Notwithstanding potential security vulnerabilities presented in the design, research reactors frequently have devices kept promptly accessible to encourage research and education. The accessibility of these sorts of hardware could be used by an authorized person to commit an unapproved activity or cause harm. This paper aims to present insights to compromise contradicting conditions in new research reactors in which both enhancing utilization and ensuring security are satisfied.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1619-1625
• /
• 2021

This paper demonstrates the possible nondestructive analysis of iron artifacts' metallurgical characteristics using neutron imaging. Ancient kingdoms of the Korean Peninsula used a direct smelting process for ore smelting and iron bloom production; however, the use of iron blooms was difficult because of their low strength and purity. For reinforcement, iron ingots were produced through refining and forge welding, which then underwent various processes to create different iron goods. To demonstrate the potential analysis using neutron imaging, while ensuring artifacts' safety, a sand iron ingot (SI-I) produced using ancient traditional iron making techniques and a sand iron knife (SI-K) made of SI-I were selected. SI-I was cut into 9 cm², whereas the entirety of SI-K was preserved for analysis. SI-I was found to have an average grain size of 3 ㎛, with observed α-Fe (ferrite) and pearlite with a body-centered cubic (BCC) lattice structure. SI-K had a grain size of 1-3 ㎛, α-Ferrite on its backside, and martensite with a body-centered tetragonal (BCT) structure on its blade. Results show that the sample's metallurgical characteristics can be identified through neutron imaging only, without losing any part of the valuable artifacts, indicating applicability to cultural artifacts requiring complete preservation.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.1013-1019
• /
• 2021

This paper reviewed the temporal changes in the public opinions on nuclear energy in Korea with a big data analysis of nuclear energy-related articles and their comments posted on the portal site NAVER. All articles that included at least one of "nuclear energy," "nuclear power plant (NPP)," "nuclear power phase-out," or "anti-nuclear" in their titles or main text were extracted from those posted on NAVER in January 2010-December 2019. First, we performed annual word frequency analysis to identify what words had appeared most frequently in the articles. For that period, the most frequent words were "NPP," "nuclear energy," and "energy." In addition, "safety" has remained in the upper ranks since the Fukushima NPP accident. Then, we performed sentiment analysis of the pre-processed articles. The sentiment analysis showed that positive-tone articles have been reported more frequently than negativetone over the entire analysis period. Last, we performed sentiment analysis of the comments on the articles to examine the public's intention regarding nuclear issues. The analysis showed that the number of negative comments to articles each month-irrespective of positive or negative tone-was always larger than that of positive comments over the entire analysis period.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.3966-3978
• /
• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1142-1153
• /
• 2019

This paper presents the effect of fusion procedure on the fusion performance of the thermal butt fusion in the safety class III buried HDPE piping per various tests performed, including high speed tensile impact, free bend, blunt notched tensile, notched creep, and PENT tests. The suitability of fusion joints and qualification procedures was evaluated by comparing test results from the base material and buttfusion joints. From the notched tensile test result, it was found that the fused joints have much lower toughness than the base material. It was also identified that the notched tensile test is more desirable than the high speed tensile impact and free bend tests presented in the ASME Code Case N-755-3 as a fusion qualification test method. In addition, with regard to the single low-pressure fusion joint performances, the procedure given by the ISO 21307 was determined to be better that the one specified in the Code Case N-755-3.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.908-914
• /
• 2019

Fusion power shutdown system (FPSS) is a safety system to stop plasma in case of accidents or incidents. The gas injection system for the FPSS presented in this work is designed to research the flow development in a closed system. As the efficiency of the system is a crucial property, plenty of experiments are executed to get optimum parameters. In this system, the flow is driven by the pressure difference between a gas storage tank and a vacuum vessel with a source pressure. The idea is based on a constant volume system without extra source gases to guarantee rapid response and high throughput. Among them, valves and gas species are studied because their properties could influence the velocity of the fluid field. Then source pressures and volumes are emphasized to investigate the volume flow rate of the injection. The source pressure has a considerable effect on the injected volume. From the data, proper parameters are extracted to achieve the best performance of the FPSS. Finally, experimental results are used as a quantitative benchmark for simulations which can add our understanding of the inner gas flow in the pipeline. In generally, there is a good consistency and the obtained correlations will be applied in further study and design for the FPSS.

• Korea Journal of Hospital Management
• /
• v.23 no.4
• /
• pp.26-39
• /
• 2018

Purposes: The purpose of this study is to analyze the issue of interest in patient medical service of small and medium hospitals using big data. Methods: The method of this study was implemented by data mining and social network using SNS big data. The analysis tool were extracted key keywords and analyzed correlation by using Textom, Ucinet6 and NetDraw program. Findings: In the results of frequency, the network-centered and closeness centrality analysis, It was shown that the government center is interested in the major explanations and evaluations of the technology, information, security, safety, cost and problems of small and medium hospitals, coping with infections, and actual involvement in bank settlement. And, were extracted care for disabilities such as pediatrics, dentistry, obstetrics and gynecology, dementia, nursing, the elderly, and rehabilitation. Practical Implications: Future studies will be more useful if analyzed the needs of customers for medical services in the metropolitan area and provinces may be different in the small and medium hospitals to be studied, further classification studies.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1041-1046
• /
• 2019

In terms of long-term safety for radioactive waste disposal, the anionic iodide (I-129) with a long half-life ($1.6{\times}10^6yr$) is of a critical importance because this radionuclide migrates in geological media with limited interactions. Various studies have been performed to retard the iodide migration. Recently, some minerals that are likely generated from waste container corrosion, have been suggested to have a considerable chemical interaction with iodide. In this study, chalcocite and mackinawite were selected as candidate minerals for underground corrosion materials, and an iodide sorption experiment were carried out. The experiment was performed under anoxic and alkaline conditions and the pH effects on the iodide sorption were investigated in the range of pH 8 to 12. The results showed that both minerals demonstrated a noticeable sorption capacity on iodide, and the distribution coefficient ($K_d$) decreased as the pH increased in the experimental condition. In addition, when the alkalinity increased higher than a pH of 12, the sorption capacity of both minerals decreased dramatically, likely due to the competition of hydroxy ions with the iodide. This result confirmed that chalcocite was an especially good sorbing media for iodide under alkaline conditions with a pH value of less than 12.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.922-929
• /
• 2019

This paper presents the drop performance test of the control rod assembly which is one of the main components strongly related to the safety of the prototype generation IV sodium-cooled fast reactor. To investigate the drop performance, a real-sized control rod assembly that was recently modified based on the drop analysis results was newly fabricated, and several free drop tests under different flow rate conditions were carried out. Then the results were compared with those obtained from the previous tests conducted on the conceptually designed control rod assembly to demonstrate the improvement in performance. Moreover, the drop performance tests under several types and magnitudes of seismic loadings were also conducted to investigate the effect of the seismic loading on the drop performance of the modified control rod assembly. The results showed that the effects of the type and magnitude of the seismic loading on the drop performance of the modified control rod assembly were not significant. Also, the drop time requirement was successfully satisfied, even under the seismic loading conditions.