권호 표지

Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) (방사성폐기물학회지)

Korean Radioactive Waste Society

  • Quarterly
  • /
  • 1738-1894(pISSN)
  • /
  • 2288-5471(eISSN)

Domain

  • Nuclear Power ＞ Nuclear Fuel Cycle/Radioactive Waste Management

Aim & Scope

Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT) is a quarterly journal publishing peer-reviewed papers in various research areas associated with the nuclear fuel cycle and radioactive waste management. The journal is managed by the Korean Radioactive Waste Society. JNFCWT provides a forum for all stakeholders in the field of nuclear fuel cycle and radioactive waste management aiming at sharing and disseminating state of the art technical information, lessons learned from industrial experiences, and discussions on current socio-political issues. JNFCWT publishes the results of original research in the format of review articles, commentaries, and technical notes. From Volume 1 (2003) to Volume 11 (2013), this journal was published as the Journal of the Korean Radioactive Waste Society. It was renamed to the Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT) in 2013. As of 2019, the journal is listed in the SCOPUS of Elsevier. JNFCWT covers the following fields related to the nuclear fuel cycle and radioactive waste management as follows: 1. Policy, Safety, and Regulation of Nuclear Fuel Cycle / Non-proliferation, Safeguards, and Nuclear Security 2. Treatment, Transportation, and Storage of Spent Nuclear Fuel / Molten Salt Chemistry and Technology 3. High-Level Waste Immobilization and Disposal Technologies 4. Low- and Intermediate-Level Radioactive Waste Management 5. Decommissioning and Environmental Remediation 6. Radiation Protection, Shielding, and Environment 7. Radiochemistry and Nuclear Chemistry

http://www.krs.or.kr KSCI KCI SCOPUS

Volume 22 Issue 2

  1. Minsoo Lee;Jin-Seop Kim;Min-Seop Kim;Seok Yoon 105
    https://doi.org/10.7733/jnfcwt.2024.011 PDF
    In this study, we employed a small-scale experiment to demonstrate the introduction of a thin copper heat dissipation plate into a bentonite buffer layer of an engineered barrier system. This experiment designed for spent nuclear fuel disposal can effectively reduce the maximum temperature of the bentonite buffer layer, and ultimately, make it possible to reduce the area of the disposal site. For the experiment, a small-scale engineered barrier system with a copper heat dissipation plate was designed and manufactured. the thickness of the cylindrical buffer was about 2 cm, which was about 1/20 of KAERI Repository System (KRS). At a power supply of 250 W, the maximum buffer temperature reduced to a mere 1.8℃ when the thin copper plate was introduced. However, the maximum surface temperature reduced to a remarkable 9.1℃, when a U-collar copper plate was introduced, which had a good contact with the other barrier layers. Consequently, we conclude that the introduction of the thin copper plate into the engineered barrier system for spent nuclear fuel disposal can effectively reduce the maximum buffer temperature in high-level radioactive waste disposal repositories.
  2. Taehyung Na;Yeji Kim;Donghee Lee;Taehyeon Kim;Sunghwan Chung 117
    https://doi.org/10.7733/jnfcwt.2024.013 PDF
    The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.
  3. Sohee Cha;Kwangheon Park 129
    https://doi.org/10.7733/jnfcwt.2024.016 PDF
    To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of 137Cs or measure based on the energy peak ratio of Cs isotopes such as 134Cs/137Cs and 154Eu/137Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming "Direct hit" and "Penetration with attenuation" situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.
  4. JaeHoon Lim;Byoungsun Park;Jongmin Lim;Yun-Young Yang;Sung-Hyo Lee;Sang Soon Cho 139
    https://doi.org/10.7733/jnfcwt.2024.018 PDF
    Concrete structures must maintain their shielding abilities and structural integrity over extended operational periods. Despite the widespread use of dry storage systems for spent nuclear fuel, research on the properties of deteriorated concrete and their impact on structural performance remains limited. To address this significant research gap, static and dynamic material testing was conducted on concrete specimens carefully extracted from the outer wall of the High-flux Advanced Neutron Application ReactOr (HANARO), constructed approximately 30 years ago. Despite its age, the results reveal that the concrete maintains its structural integrity impressively well, with static compression tests indicating an average compressive strength exceeding the original design standards. Further dynamic property testing using advanced high-speed material test equipment supported these findings, showing the consistency of dynamic increase factors with those reported in previous studies. These results highlight the importance of monitoring and assessing concrete structures in nuclear facilities for long-term safety and reliability.
  5. Seungil Ha;Dalhyeon Ryu;Giyoon Kim;Myungsoo Kim 145
    https://doi.org/10.7733/jnfcwt.2024.019 PDF
    Undeclared nuclear activities are challenging given the lack of information from the sites involved in such activities. Wide-area environmental sampling (WAES) can be an effective method to detect undeclared nuclear activities. However, it is crucial to address the potential risks during the WAES, including sample tampering or extortions. Therefore, tracking and monitoring of various on-site data is imperative to accurately interpret the status of samples and workers throughout the WAES process. 'Environmental and Geographical Data Transfer (EGDT)' was developed for the real-time monitoring of integrated on-site data. EGDT module is equipped with various sensors and can be attached to a worker's uniform or a sample storage box. This study demonstrated the technical effectiveness of EGDT by exploring three experimental methodologies for feasibility assessment. Compared to the Normal Operation case, the inference of the Sample Extortion case was predominantly based on changes in lux and dose rate. The inference of the Out-of-Work-Area case primarily relied on changes in dose rate and acceleration. Finally, the preliminary evaluation of the performance of the developed prototype was conducted, and a foundation was established for enhancing the application in the WAES process.
  6. Hyun-Kyu Lee;Byung-Moon Jun;Tack-Jin Kim;Sungbin Park;Seonggyu Choi;Jun-Young Jung;Hee-Chul Eun 159
    https://doi.org/10.7733/jnfcwt.2024.020 PDF
    This study aimed to develop an efficient recycling process for wastewater generated from soil-washing used to remediate uranium (U(VI))-contaminated soil. Under acidic conditions, U(VI) ions leached from the soil were precipitated and separated through neutralization using hydrazine (N2H4). N2H4, employed as a pH adjuster, was decomposed into nitrogen gas (N2), water (H2O), and hydrogen ions (H+) by hydrogen peroxide (H2O2). The residual N2H4 was precipitated when the pH was adjusted using sulfuric acid (H2SO4) to recycle the wastewater in the soil-washing process. This purified wastewater was reused in the soil-washing process for a total of ten cycles. The results confirmed that the soil-washing performance for U(VI)-contaminated soil was maintained when using recycled wastewater. All in all, this study proposes an efficient recycling process for wastewater generated during the remediation of U(VI)-contaminated soil.
  7. Young Hwan Hwang;Seong-Sik Shin;Sunghoon Hong;Jung-Kwon Son;Cheon-Woo Kim 173
    https://doi.org/10.7733/jnfcwt.2024.022 PDF
    Vitrification, one of the most promising solidification processes for various materials, has been applied to radioactive waste to improve its disposal stability and reduce its volume. Because the thermal decomposition of dry active waste (DAW) significantly reduces its volume, the volume reduction factor of DAW vitrification is high. The KHNP developed the optimal glass composition for the vitrification of DAW. Since vitrification offers a high-volume reduction ratio, it is expected that disposal costs could be greatly reduced by the use of such technology. The DG-2 glass composition was developed to vitrify DAW. During the maintenance of nuclear power plants, metals containing paper, clothes, and wood are generated. ZrO2 and HfO2 are generally considered to be network-formers in borosilicate-based glasses. In this study, a feasibility study of vitrification for DAW that contains metal particulates is conducted to understand the applicability of this process under various conditions. The physicochemical properties are characterized to assess the applicability of candidate glass compositions.
  8. Mijin Kwon;Seho Choi;Chunhyung Cho 185
    https://doi.org/10.7733/jnfcwt.2024.012 PDF
    Disposal cover as an engineered barrier of a near-surface disposal facility for low and very low-level radioactive waste is composed of a multi-layer to isolate radioactive waste from environmental influences for the long term. To acquire a realistic forecast for the post-closure period of the disposal facility, it is essential to carry out long-term experimental research in a similar condition to the actual disposal environment. Hence, a performance test facility of the disposal cover was constructed in Gyeongju low and intermediate level radioactive waste disposal center in 2022. The constructed performance test facility has differences from the material properties presented in the design. These differences are factors that affect the prevent rainfall infiltration, which is one of the important roles of the disposal cover. Therefore, in this study, a numerical simulation of rainfall infiltration into the performance test facility was performed for the designed case and the actual constructed case. To simulate the behavior of water infiltration, the FEFLOW software based on the finite element method is used. Through the analysis of numerical simulation results, it is confirmed that the hydraulic conductivity of the material constituting the multi-layer of the disposal cover greatly influences the amount of water infiltration.
  9. Taehyung Na;Youngoh Lee;Yeji Kim;Donghee Lee;Taehyeon Kim;Kiyoung Kim;Yongdeog Kim 201
    https://doi.org/10.7733/jnfcwt.2024.014 PDF
    Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.
  10. Gha-Young Kim;Jeong-Hyun Woo;Junhyuk Jang;Yang-Il Jung;Young-Ho Lee 211
    https://doi.org/10.7733/jnfcwt.2024.015 PDF
    The growing significance of sustainable energy technologies underscores the need for safe and efficient management of spent nuclear fuels (SNFs), particularly via deep geological disposal (DGD). DGD involves the long-term isolation of SNFs from the biosphere to ensure public safety and environmental protection, necessitating materials with high corrosion resistance for DGD canisters. This study investigated the feasibility of a Cu-Ni film, fabricated via additive manufacturing (AM), as a corrosion-resistant layer for DGD canister applications. A wire-fed AM technique was used to deposit a millimeter-scale Cu-Ni film onto a carbon steel (CS) substrate. Electrochemical analyses were conducted using aerated groundwater from the KAERI underground research tunnel (KURT) as an electrolyte with an NaCl additive to characterize the oxic corrosion behavior of the Cu-Ni film. The results demonstrated that the AM-fabricated Cu-Ni film exhibited enhanced corrosion resistance (manifested as lower corrosion current density and formation of a dense passive layer) in an NaCl-supplemented groundwater solution. Extensive investigations are necessary to elucidate microstructural performance, mechanical properties, and corrosion resistance in the presence of various corroding agents to simplify the implementation of this technology for DGD canisters.
  11. Seongjin Jeong;Jinseok Han;Hyun Chul Lee 219
    https://doi.org/10.7733/jnfcwt.2024.017 PDF
    Democratic People's Republic of Korea (DPRK) has produced weapon-grade plutonium in a graphite-moderated experimental reactor at the Yongbyon nuclear facilities. The amount of plutonium produced can be estimated using the Graphite Isotope Ratio Method (GIRM), even without considering specific operational histories. However, the result depends to some degree on the operational cycle length. Moreover, an optimal cycle length can maximize the number of nuclear weapons made from the plutonium produced. For conservatism, it should be assumed that the target reactor was operated with an optimal cycle length. This study investigated the optimal cycle length using which the Calder Hall MAGNOX reactor can achieve the maximum annual production of nuclear weapons. The results show that lower enrichment fuel produced a greater number of critical plutonium spheres with a shorter optimal cycle length. Specifically, depleted uranium (0.69wt%) produced 5.561 critical plutonium spheres annually with optimal cycle lengths of 251 effective full power days. This research is crucial for understanding DPRK's potential for nuclear weapon production and highlights the importance of reactor operational strategy in maximizing the production of weapons-grade plutonium in MAGNOX reactors.
  12. Young Hwan Hwang;Sunghoon Hong;Seong-Sik Shin;Seokju Hwang;Jung-Kwon Son;Cheon-Woo Kim;Changgyu Kim;Kwang Soo Park;Taeseob Lim;Donghun Park 227
    https://doi.org/10.7733/jnfcwt.2024.021 PDF
    After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.