간행물
방사성폐기물학회지 KCI 등재 SCOPUS Journal of the Korean Radioactive Waste Society
- 발행기관 한국방사성폐기물학회
- 자료유형 학술지
- 간기 계간
- ISSN 1738-1894 (Print)2288-5471 (Online)
- 수록기간 2003 ~ 2024
- 주제분류 공학 > 공학일반 공학 분류의 다른 간행물
- 십진분류KDC 559DDC 629
권호리스트/논문검색
Volume 22 Number 3 (2024년 9월) 12건
1.
2024.09
구독 인증기관 무료, 개인회원 유료
The initial radionuclide migration quantity depends on the total amount of solubilized species. Geochemical modeling based on a thermodynamic database (TDB) has been employed to assess the solubility of radionuclides. It is necessary to evaluate whether the TDB describes the domestic repository conditions appropriately. An effective way to validate the TDB-based modeling results is through direct comparisons with experimentally measured values under the conditions of interest. Here, the solubilities of trivalent Sm, Eu, and Am were measured in synthetic KURT-DB3 groundwater (Syn- DB3) and compared with modeling results based on ThermoChimie TDB. Ln2(CO3)3·xH2O(cr) (Ln = Sm, Eu) solids were introduced into the Syn-DB3 and dissolved Sm and Eu concentrations were monitored over 223 days. X-ray diffraction analysis confirmed that the crystallinity of the solid compounds was maintained throughout the experiments. The dissolved Sm and Eu concentrations at equilibrium were close to the predicted solubilities of Sm2(CO3)3(s) and Eu2(CO3)3(s) based on the ThermoChimie TDB. The Am solubility measured under oversaturated conditions was comparable to the measured Eu concentrations, although they were measured under different experimental settings. More experimental data are needed for Am-carbonate solid systems with careful characterization of the solid phases to better evaluate Am solubility in domestic groundwater conditions.
20,000원
2.
2024.09
구독 인증기관 무료, 개인회원 유료
Pyrochemical processing and molten-salt reactors have recently garnered significant attention as they are promising options for future nuclear technologies, such as those for recycling spent nuclear fuels and the next generation of nuclear reactors. Both of these technologies require the use of high-temperature molten salt. To implement these technologies, one must understand the electrochemical behavior of fission products in molten salts, lanthanides, and actinides. In this study, a rotating-disk-electrode (RDE) measurement system for high-temperature molten salts is constructed and tested by investigating the electrochemical reactions of Sm3+ in LiCl–KCl melts. The results show that the reduction of Sm3+ presents the Levich behavior in LiCl–KCl melts. Using the RDE system, not only is the diffusion-layer thickness of Sm3+ measured in high-temperature molten salts but also various electrochemical parameters for Sm3+ in LiCl–KCl melts, including the diffusion coefficient, Tafel slope, and exchange current density, are determined.
4,000원
3.
2024.09
구독 인증기관 무료, 개인회원 유료
A procedure for minimizing the environmental burden and maximizing the efficiency of storage sites used for the final disposal of spent fuel has been proposed. In this procedure, fission products (highly mobile and producing heat) are collected, and uranium and TRU-RE (transuranium-rare earth) oxide are independently stored. The possibility and applicability of radiation measurement for monitoring the nuclear materials effectively throughout the process has been simulated and evaluated. For the simulation, the properties of the chemical processes were analyzed, the major radiation emitters were determined, and the production of nuclear materials by chemical reactions were evaluated. In each process, the content of nuclear material was changed by up to 20% to represent abnormal conditions. The results showed that the plutonium peak was matched with the change in the TRU content and the measured signal was changed linearly with respect to the content change of the plutonium. From the neutron measurement, a linear response of the TRU content variation was obtained. In addition, a logic diagram was developed for the nuclear monitoring. The integration of radiation detections is recommended for monitoring the process effectively and efficiently.
4,500원
4.
2024.09
구독 인증기관 무료, 개인회원 유료
Technetium-99 is identified as an element of interest for the safety assessment of a deep geological repository for used nuclear fuel. The sorption behavior of Tc(IV) onto MX-80 and granite in Ca-Na-Cl solutions of varying ionic strength (0.05–1 mol·kgw−1 (m)) and across a pHm range of 4–9 was studied in this paper. Sorption of Tc(IV) was found to be independent of ionic strength in the range of 0.05 to 1 m for both MX-80 and granite. Sorption of Tc(IV) on MX-80 increased with pHm from 4 to 7 and then decreased with pHm from 8 to 9. Sorption of Tc(IV) on granite gradually increased with pHm from 4 to 8 and then became almost constant or slightly decreased with pHm from 8 to 9. A 2 site protolysis non-electrostatic surface complexation and cation exchange sorption model successfully simulated sorption of Tc(IV) on MX-80 and granite. Optimized values of surface complexation constants (log K0) are proposed.
4,500원
5.
2024.09
구독 인증기관 무료, 개인회원 유료
Technetium has been identified as an element of interest for the safety assessment of a deep geological repository for used nuclear fuel. In this study, the sorption of Tc(IV) onto MX-80 bentonite, illite, and shale in ionic strength (I) 0.1–6 mol·kgw−1 (m) Na-Ca-Cl solutions at pHm = 4–9 and limestone at pHm = 5–9 was studied. Tc(IV) sorption on MX-80 increased with pHm from 4 to 6, reached the maximum at pHm = 6–7, and then gradually decreased with pHm from 7 to 9. Tc(IV) sorption on illite gradually increased with pHm from 4 to 7, and then decreased as pHm increased. The sorption properties of Tc(IV) on shale were quite similar to those on illite. Tc(IV) sorption on limestone slightly increased with pHm from 5 to 6 and then seemed to be constant at pHm = 6–9. Tc(IV) sorption on all four solids was independent of ionic strength (0.1–6 m). The 2 site protolysis non-electrostatic surface complexation and cation exchange model successfully simulated the sorption of Tc(IV) onto MX-80 and illite and the optimized values of surface complexation constants were estimated.
4,500원
6.
2024.09
구독 인증기관 무료, 개인회원 유료
Hydride analysis is required to assess the mechanical integrity of spent nuclear fuel cladding. Image segmentation, which is a hydride analysis method, is a technique that can analyze the orientation and distribution of hydrides in cladding images of spent nuclear fuels. However, the segmentation results varied according to the image preprocessing. Inaccurate segmentation results can make hydride difficult to analyze. This study aims to analyze the segmentation performance of the Otsu algorithm according to the morphological operations of cladding images. Morphological operations were applied to four different cladding images, and segmentation performance was quantitatively compared using a histogram, betweenclass variance, and radial hydride fraction. As a result, this study found that morphological operations can induce errors in cladding images and that appropriate combinations of morphological operations can maximize segmentation performance. This study emphasizes the importance of image preprocessing methods, suggesting that they can enhance the accuracy of hydride analysis. These findings are expected to contribute to the advancements in integrity assessment of spent nuclear fuel cladding.
4,200원
7.
2024.09
구독 인증기관 무료, 개인회원 유료
Se sorption onto Ca-type montmorillonite purified from Bentonil-WRK—a new research bentonite introduced by Korea Atomic Energy Research Institute—was examined under ambient conditions (pH 4−9, pe 7−9, I = 0.01 M CaCl2, and T = 25°C). Se(IV) was identified as the oxidation state responsible for weak sorption (Kd < 22 L∙kg−1) by forming surface complexes with edge functional groups of the montmorillonite. Thermodynamic modeling, considering reaction mechanisms of outer-sphere complexation (≡AlOH2 + + HSeO3 − ⇌ ≡AlOH3SeO3, log K = 0.50 ± 0.21), inner-sphere complexation (2≡AlOH + H2SeO3(aq) ⇌ (≡Al)2SeO3 + 2H2O(l), log K = 7.89 ± 0.51), and Ca2+-involved ternary complexation (≡AlOH + Ca2+ + SeO3 2− ⇌ ≡AlOHCaSeO3, log K = 7.69 ± 0.28) between selenite and aluminol sites of montmorillonite, acceptably reproduced the batch sorption data. Outer- and inner-sphere complexes are predominant Se(IV) forms sorbed in acidic (pH ≈ 4) and near-acidic (pH ≈ 6) regions, respectively, whereas ternary complexation accounts for Se(IV) sorption at neutral pHs under the ambient conditions. The experimental and modeling data generally extend a material-specific sorption database of Bentonil-WRK, which is essential for assessing its radionuclide retention performance as a buffer candidate of deep geological disposal system for high-level radioactive waste.
4,300원
8.
2024.09
구독 인증기관 무료, 개인회원 유료
Interim dry cask storage systems comprising AISI 304 or 316 stainless steel canisters have become critical for the storage of spent nuclear fuel from light water reactors in the Republic of Korea. However, the combination of microstructural sensitization, residual tensile stress, and corrosive environments can induce chloride-induced stress corrosion cracking (CISCC) for stainless steel canisters. Suppressing one or more of these three variables can effectively mitigate CISCC initiation or propagation. Surface-modification technologies, such as surface peening and burnishing, focus on relieving residual tensile stress by introducing compressive stress to near-surface regions of materials. Overlay coating methods such as cold spray can serve as a barrier between the environment and the canister, while also inducing compressive stress similar to surface peening. This approach can both mitigate CISCC initiation and facilitate CISCC repair. Surface-painting methods can also be used to isolate materials from external corrosive environments. However, environmental variables, such as relative humidity, composition of surface deposits, and pH can affect the CISCC behavior. Therefore, in addition to research on surface modification and coating technologies, site-specific environmental investigations of various nuclear power plants are required.
4,600원
9.
2024.09
구독 인증기관 무료, 개인회원 유료
The spent nuclear fuel, combusted and released in the nuclear power plant, is stored in the spent fuel pool (SFP) located in the fuel buildings interconnected with the reactors. In Korea, spent fuel has been stored exclusively in SFPs, prompting initiatives to expand storage capacity by either installing additional SFPs or replacing them with high-density spent fuel storage racks. The installation of these fuel racks necessitates obtaining a regulatory license contingent upon ensuring safe fuel handling and storage systems. Regulatory agencies mandate the formulation of various postulated accident scenarios and assessments covering criticality, shielding, thermal behavior, and structural integrity to ensure safe fuel handling and storage systems. This study describes an evaluation method for assessing the structural damage to storage racks resulting from fuel dropping as a part of the functional safety evaluation of these racks. A scenario was envisaged wherein fuel was dropped onto the base plates of the upper and lower sections of the storage racks, and the impact load was analyzed using the ABAQUS/Explicit program. The evaluation results revealed localized plastic deformation but affirmed the structural integrity and safety of the storage racks.
4,000원
10.
2024.09
구독 인증기관 무료, 개인회원 유료
In Korea, two types of spent nuclear fuels (SNFs) are generated, pressurized light water reactor type (PWR) and pressurized heavy water reactor type (PHWR; CANDU), that differ greatly in size, decay heat, and radioactive characteristics. Technology development for the disposal of SNFs has mainly focused on PWR SNFs that are large in size and have extremely high decay heat and radioactivity. However, CANDU SNFs should be considered differently from PWR SNFs in deep geological disposal systems because their characteristics significantly differ from those of PWR SNFs in terms of their dimensions, number of SNF bundles, and handling systems in nuclear power plant sites. In this paper, after reviewing the status of the CANDU SNF disposal concept by Canada and Korea, concepts related to the direct geological disposal of CANDU SNFs were described, and two concepts were proposed based on the results of the development. The engineered barrier systems developed using these two concepts were comparatively analyzed in terms of disposal safety, disposal efficiency, and technical maturity. Based on the results of the comparative analyses, a vertical-type emplacement disposal concept was determined as a reference concept for the deep geological disposal of CANDU SNFs.
4,900원
11.
2024.09
구독 인증기관 무료, 개인회원 유료
Jaeok Park, Tae Young Kong, Seongjun Kim, Jinho Son, Changju Song, Jiung Kim, Seungho Jo, Hee Geun Kim
Kori Unit 1, the first commercial nuclear power plant (NPP) in Korea, was permanently shut down in 2017 and was scheduled for decommissioning. Various programs must be planned early in the decommissioning process to safely decommission NPPs. Radiological characterization is a key program in decommissioning and should be a high priority. Radiological characterization involves determining the decommissioning technology to be applied to a nuclear facility by identifying the radiation sources and radioactive contaminants present within the facility and assessing the extent and nature of the radioactive contaminants to be removed from the facility. This study introduces the regulatory requirements, procedures, and implementation methods for radiological characterization and proposes a methodology to link the results of radiological characterizations for each stage. To link radiological characteristics, this study proposes to conduct radiological characterization in the decommissioning phase to verify the results of radiological characterization in the transitional phase of decommissioning NPPs. This enables significantly reducing the scope and content of radiological characterization that must be performed in the decommissioning phase and maintaining the connection with the previous phase.
4,600원
12.
2024.09
구독 인증기관 무료, 개인회원 유료
The concrete silo dry storage system, which has been in operation at the Wolsong NPP site since 1992, consists of a concrete structure, a steel liner plate in the inner space, and a fuel basket. The silo system’s concrete structure must maintain structural integrity as well as adequate radiation shielding performance against the high radioactivity of spent nuclear fuel stored inside the storage system. The concrete structure is directly exposed to the external climatic environment in the storage facility and can be expected to deteriorate over time owing to the heat of spent nuclear fuel, as well as particularly cracks in the concrete structure. These cracks may reduce the radiation shielding performance of the concrete structure, potentially exceeding the silo system’s allowable radiation dose rate limits. For specimens with the same composition and physical properties as silo’s concrete structures, cracks were forcibly generated and then irradiated to measure the change in radiation dose rate to examine the effect of cracks in concrete structures on radiation shielding performance, and in the current state, the silo system maintains radiation shielding performance.
4,000원
