Pub Date : 2024-08-06DOI: 10.1016/j.net.2024.08.011
Mumtaz Ali, Ahmed Samour, Suhaib Ahmed Soomro, Waqar Khalid, Turgut Tursoy
Nuclear energy is considered an effective means of enhancing environmental sustainability. Considering this point, this study aims to explore the impact of nuclear energy, financial globalization, technological innovation, and economic growth on ecological sustainability in the top-10 nuclear energy-consuming economies from 1995 to 2020. The load capacity factor is used as a novel proxy for ecological sustainability, explaining how human actions affect ecological sustainability and how nature compensates for human-induced damage. The study employs a novel non-parametric MMQR approach to obtain coefficients across heterogeneous quantiles. The MMQR estimation findings indicate that: (i) nuclear and renewable energy consumption and financial globalization promote environmental sustainability by increasing the LCF; (ii) economic growth degrades ecological sustainability by decreasing the LCF; and (iii) the results from Granger causality suggest a causal link among economic growth, technological innovation, nuclear energy, and LCF. The study recommends that the governments of the top-10 nuclear energy-consuming countries facilitate more investment in green technologies and green energy to achieve environmental sustainability.
{"title":"A step towards a sustainable environment in top-10 nuclear energy consumer countries: The role of financial globalization and nuclear energy","authors":"Mumtaz Ali, Ahmed Samour, Suhaib Ahmed Soomro, Waqar Khalid, Turgut Tursoy","doi":"10.1016/j.net.2024.08.011","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.011","url":null,"abstract":"Nuclear energy is considered an effective means of enhancing environmental sustainability. Considering this point, this study aims to explore the impact of nuclear energy, financial globalization, technological innovation, and economic growth on ecological sustainability in the top-10 nuclear energy-consuming economies from 1995 to 2020. The load capacity factor is used as a novel proxy for ecological sustainability, explaining how human actions affect ecological sustainability and how nature compensates for human-induced damage. The study employs a novel non-parametric MMQR approach to obtain coefficients across heterogeneous quantiles. The MMQR estimation findings indicate that: (i) nuclear and renewable energy consumption and financial globalization promote environmental sustainability by increasing the LCF; (ii) economic growth degrades ecological sustainability by decreasing the LCF; and (iii) the results from Granger causality suggest a causal link among economic growth, technological innovation, nuclear energy, and LCF. The study recommends that the governments of the top-10 nuclear energy-consuming countries facilitate more investment in green technologies and green energy to achieve environmental sustainability.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207268","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-05DOI: 10.1016/j.net.2024.08.009
Hyun Chul Lee, Jungjin Bang, Kiwan Jang, Min Beom Heo, Doo Yong Lee, Daewoo Kim, Chang-Ha Lee
After the severe accident at the Fukushima nuclear power plant, there has been heightened interest in the development of severe accident countermeasures. As part of these efforts, a passive filtration system capable of passive operation during power loss is under development.
{"title":"Exothermic characteristics evaluation of the Pt coated α-Al2O3 catalyst in bed for passive airborne radioactive material reduction system under forced flow conditions","authors":"Hyun Chul Lee, Jungjin Bang, Kiwan Jang, Min Beom Heo, Doo Yong Lee, Daewoo Kim, Chang-Ha Lee","doi":"10.1016/j.net.2024.08.009","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.009","url":null,"abstract":"After the severe accident at the Fukushima nuclear power plant, there has been heightened interest in the development of severe accident countermeasures. As part of these efforts, a passive filtration system capable of passive operation during power loss is under development.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207270","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-05DOI: 10.1016/j.net.2024.08.010
Hae Min Park, Jong Hyuk Lee, Chiwoong Choi, Kwi-Seok Ha, Byung Hyun You, Jaeseok Heo, Kyung Doo Kim, Sung Won Bae, Seung Wook Lee, Dong Hyuk Lee, Sang Ik Lee, Chan Eok Park, Bub Dong Chung, Kwang Won Seul
Recent safety issues such as cladding oxidation and fuel fragmentation, relocation and dispersal (FFRD) make the loss-of-coolant accident (LOCA) acceptance criteria more difficult to be satisfied. To obtain the adequate safety margin and more economical use of the nuclear power plant, re-classification of LOCAs for Korean operating PWRs is now under consideration to exclude the large break LOCA (LBLOCA) from the design basis accidents (DBAs). Therefore, the intermediate break LOCA (IBLOCA) might become the limiting break size of concern in the LOCA analysis. To accomplish this reform of LOCA classification, an extensive understanding of IBLOCA is crucial, and the applicability of safety analysis code should be confirmed. For this purpose, IBLOCA PIRT was developed. The PIRT panel was organized and the general process of the PIRT development established by Wilson and Boyack (1998) was adopted to develop the IBLOCA PIRT. Based on IBLOCA analyses using the safety and performance analyzing code (SPACE), the PIRT panel defined the temporal phases, systems, components and possible phenomena during an IBLOCA. For all possible phenomena, the relative importance and the knowledge level were determined for each component and each phase via discussions among PIRT panel members. For phenomena having relatively high importance, a code validation matrix was also developed. The results of the IBLOCA PIRT will be used to improve the SPACE code for IBLOCA application and resolve future regulatory issues.
{"title":"Development of a phenomena identification and ranking table (PIRT) for intermediate break loss-of-coolant accident in PWRs","authors":"Hae Min Park, Jong Hyuk Lee, Chiwoong Choi, Kwi-Seok Ha, Byung Hyun You, Jaeseok Heo, Kyung Doo Kim, Sung Won Bae, Seung Wook Lee, Dong Hyuk Lee, Sang Ik Lee, Chan Eok Park, Bub Dong Chung, Kwang Won Seul","doi":"10.1016/j.net.2024.08.010","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.010","url":null,"abstract":"Recent safety issues such as cladding oxidation and fuel fragmentation, relocation and dispersal (FFRD) make the loss-of-coolant accident (LOCA) acceptance criteria more difficult to be satisfied. To obtain the adequate safety margin and more economical use of the nuclear power plant, re-classification of LOCAs for Korean operating PWRs is now under consideration to exclude the large break LOCA (LBLOCA) from the design basis accidents (DBAs). Therefore, the intermediate break LOCA (IBLOCA) might become the limiting break size of concern in the LOCA analysis. To accomplish this reform of LOCA classification, an extensive understanding of IBLOCA is crucial, and the applicability of safety analysis code should be confirmed. For this purpose, IBLOCA PIRT was developed. The PIRT panel was organized and the general process of the PIRT development established by Wilson and Boyack (1998) was adopted to develop the IBLOCA PIRT. Based on IBLOCA analyses using the safety and performance analyzing code (SPACE), the PIRT panel defined the temporal phases, systems, components and possible phenomena during an IBLOCA. For all possible phenomena, the relative importance and the knowledge level were determined for each component and each phase via discussions among PIRT panel members. For phenomena having relatively high importance, a code validation matrix was also developed. The results of the IBLOCA PIRT will be used to improve the SPACE code for IBLOCA application and resolve future regulatory issues.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207275","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-05DOI: 10.1016/j.net.2024.08.007
Qing Zhao, Shuo Meng, Longcheng Liu
In this work, a new Time Domain Random Walk (TDRW) algorithm is proposed to estimate the tracer distribution profile within the rock matrix. The development of the new algorithm stems from the statistical properties of the analytical solution to a single fracture-matrix system, in which the particle position at a certain time is calculated and recorded. With the position of each particle determined, the resulting distribution will then provide an estimate of the tracer distribution profile directly. In addition, the newly developed algorithm can readily be extended to a case of more complicated fracture-matrix system, in which an arbitrary injection boundary condition may also be used. To verify the accuracy and applicability of the new algorithm, three benchmark simulations are made, in which the results of different approaches are found to be identical. Nevertheless, the new algorithm has a higher computational efficiency, due to its lower calculation demand.
{"title":"Tracer transport in fractured porous media: Distribution of tracer concentration within the rock matrix and the implementation of time domain random walk algorithm","authors":"Qing Zhao, Shuo Meng, Longcheng Liu","doi":"10.1016/j.net.2024.08.007","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.007","url":null,"abstract":"In this work, a new Time Domain Random Walk (TDRW) algorithm is proposed to estimate the tracer distribution profile within the rock matrix. The development of the new algorithm stems from the statistical properties of the analytical solution to a single fracture-matrix system, in which the particle position at a certain time is calculated and recorded. With the position of each particle determined, the resulting distribution will then provide an estimate of the tracer distribution profile directly. In addition, the newly developed algorithm can readily be extended to a case of more complicated fracture-matrix system, in which an arbitrary injection boundary condition may also be used. To verify the accuracy and applicability of the new algorithm, three benchmark simulations are made, in which the results of different approaches are found to be identical. Nevertheless, the new algorithm has a higher computational efficiency, due to its lower calculation demand.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207271","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-05DOI: 10.1016/j.net.2024.08.006
Tao Zhou, Yong Zhu, Shengnan Tang
Optimizing the hydraulic components of nuclear main pump (NMP) and conducting performance verification is crucial. Due to the large size of the real NMP, the strict requirements of the operation and the high test-cost, there are many difficulties in the real test. The mixed flow NMP is taken as the research object, and the CAP1400 NMP is selected as the prototype pump (PP). The model pumps (MPs) with varying scales are established based on the similarity conversion algorithm (SCA). Then, the influence of different scales on the hydraulic performance and internal flow field is investigated and compared. It is demonstrated that the predicted value of head is 4 % higher than the design value at the design operating point, and the maximum efficiency point is close to the design operating point. In the range of full flow conditions, the head, hydraulic efficiency, impeller efficiency, guide vane energy loss, internal flow field, and vorticity distribution of PP and MPs are basically consistent with the trend of flow rate variations. The PP and MPs conform to the SCA. The hydraulic design and performance optimization of NMP are achieved by using the model proportional scaling approach.
{"title":"Scale effect for hydraulic model of a mixed flow nuclear main pump","authors":"Tao Zhou, Yong Zhu, Shengnan Tang","doi":"10.1016/j.net.2024.08.006","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.006","url":null,"abstract":"Optimizing the hydraulic components of nuclear main pump (NMP) and conducting performance verification is crucial. Due to the large size of the real NMP, the strict requirements of the operation and the high test-cost, there are many difficulties in the real test. The mixed flow NMP is taken as the research object, and the CAP1400 NMP is selected as the prototype pump (PP). The model pumps (MPs) with varying scales are established based on the similarity conversion algorithm (SCA). Then, the influence of different scales on the hydraulic performance and internal flow field is investigated and compared. It is demonstrated that the predicted value of head is 4 % higher than the design value at the design operating point, and the maximum efficiency point is close to the design operating point. In the range of full flow conditions, the head, hydraulic efficiency, impeller efficiency, guide vane energy loss, internal flow field, and vorticity distribution of PP and MPs are basically consistent with the trend of flow rate variations. The PP and MPs conform to the SCA. The hydraulic design and performance optimization of NMP are achieved by using the model proportional scaling approach.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207272","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-03DOI: 10.1016/j.net.2024.08.004
Maryam Al Huwayz, Aljawhara H. Almuqrin, F.F. Alharbi, M.I. Sayyed, B. Albarzan
This research focuses on the preparation of a new glass system designed specifically for applications in radiation shielding materials. These glasses are based on the general formula (56-x)BO–10TiO–8BaO–27ZnO-(x-1)NdO, where x takes the values of 2, 4, 6 and 8 mol%. For the examination of the designed glasses' radiation attenuation performance, Phy-X software was used, which is a useful approach for predicting the linear attenuation coefficient (LAC), the half value layer (HVL), and effective atomic number. The LAC decreases from 1.489 cm to 0.551 cm for the glass with x = 1 mol%, while the glass with 7 mol% NdO saw a decrease in the LAC from 2.483 cm to 0.718 cm. Introducing NdO increases the glasses' LAC, suggesting enhanced radiation shielding performance. Also, NdO addition influences the HVL within the glasses, with higher content reducing the HVL. At 0.122 MeV, the HVL and tenth value layer (TVL) are 0.456 and 1.546 cm, respectively. At 0.245 MeV, the TVL is about 3.32 times higher than the HVL. The lowest mean free path (MFP) is found at 0.122 MeV, which varies between 0.672 cm for Nd1 and 0.403 cm for Nd4.
{"title":"Unveiling the potential of Nd2O3 in optimizing the radiation shielding performance of B2O3–TiO2–BaO–ZnO-Nd2O3 glasses","authors":"Maryam Al Huwayz, Aljawhara H. Almuqrin, F.F. Alharbi, M.I. Sayyed, B. Albarzan","doi":"10.1016/j.net.2024.08.004","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.004","url":null,"abstract":"This research focuses on the preparation of a new glass system designed specifically for applications in radiation shielding materials. These glasses are based on the general formula (56-x)BO–10TiO–8BaO–27ZnO-(x-1)NdO, where x takes the values of 2, 4, 6 and 8 mol%. For the examination of the designed glasses' radiation attenuation performance, Phy-X software was used, which is a useful approach for predicting the linear attenuation coefficient (LAC), the half value layer (HVL), and effective atomic number. The LAC decreases from 1.489 cm to 0.551 cm for the glass with x = 1 mol%, while the glass with 7 mol% NdO saw a decrease in the LAC from 2.483 cm to 0.718 cm. Introducing NdO increases the glasses' LAC, suggesting enhanced radiation shielding performance. Also, NdO addition influences the HVL within the glasses, with higher content reducing the HVL. At 0.122 MeV, the HVL and tenth value layer (TVL) are 0.456 and 1.546 cm, respectively. At 0.245 MeV, the TVL is about 3.32 times higher than the HVL. The lowest mean free path (MFP) is found at 0.122 MeV, which varies between 0.672 cm for Nd1 and 0.403 cm for Nd4.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207273","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-03DOI: 10.1016/j.net.2024.08.003
Yun Soo Lim, Dong Jin Kim, Sung Woo Kim, Seong Sik Hwang, Sung Hwan Cho, Hong Pyo Kim
The internal oxidation (IO) and preferential intergranular oxidation (PIO) behavior of Alloy 600 depending on the dissolved hydrogen (DH) content and the IG Cr carbide in hydrogenated primary water were characterized in detail using analytical electron microscopy techniques. The oxidation layer was unstable when the DH concentration was such that Ni was in the vicinity of Ni/NiO equilibrium and it could easily be peeled off. Hence, the grain boundaries of the bare metal were attacked. PIO occurred and Cr-rich oxide identified as CrO was formed at the oxidized grain boundary. NiO emerged when the DH concentration was such that Ni was in an oxidizing state, whereas Ni enrichment occurred inside the oxidized grain boundary when the DH concentration was such that Ni was in a reducing state with respective to Ni/NiO equilibrium. The IG Cr carbide strongly affected the PIO behavior by means of the consumption of oxygen penetrating into the grain boundary. The depth of the IO layer decreased as the DH concentration increased. The different oxidation behaviors depending on the DH content and IG Cr carbide are believed to affect the PWSCC resistance of Alloy 600 significantly.
{"title":"Influence of dissolved hydrogen and IG Cr carbide on the oxidation behavior of Alloy 600 in hydrogenated primary water","authors":"Yun Soo Lim, Dong Jin Kim, Sung Woo Kim, Seong Sik Hwang, Sung Hwan Cho, Hong Pyo Kim","doi":"10.1016/j.net.2024.08.003","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.003","url":null,"abstract":"The internal oxidation (IO) and preferential intergranular oxidation (PIO) behavior of Alloy 600 depending on the dissolved hydrogen (DH) content and the IG Cr carbide in hydrogenated primary water were characterized in detail using analytical electron microscopy techniques. The oxidation layer was unstable when the DH concentration was such that Ni was in the vicinity of Ni/NiO equilibrium and it could easily be peeled off. Hence, the grain boundaries of the bare metal were attacked. PIO occurred and Cr-rich oxide identified as CrO was formed at the oxidized grain boundary. NiO emerged when the DH concentration was such that Ni was in an oxidizing state, whereas Ni enrichment occurred inside the oxidized grain boundary when the DH concentration was such that Ni was in a reducing state with respective to Ni/NiO equilibrium. The IG Cr carbide strongly affected the PIO behavior by means of the consumption of oxygen penetrating into the grain boundary. The depth of the IO layer decreased as the DH concentration increased. The different oxidation behaviors depending on the DH content and IG Cr carbide are believed to affect the PWSCC resistance of Alloy 600 significantly.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141940106","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-03DOI: 10.1016/j.net.2024.08.001
Yin Qiu, Ming Li, Xiaolong Liu, Qinghua Ren, Tao Lin, Jian Liu
The TF coil of HL-3 consists of 20 D-shaped demountable bundles, and all their L-shaped center sections are bonded together as the center-post. To increase its overall rigidity to resist torsional loads and avoid de-bonding failure, the center-post is wrapped by an insulation cylinder with enough prestress. Additionally, the insulation cylinder needs to be of good dimensional precision because the CS coil is winded around it. This paper describes the results of the technical studies carried out during the development of the insulation cylinder, including material selection, parameter calculation, enlacing tests and curing tests.
HL-3 的 TF 线圈由 20 个 D 型可拆卸线束组成，其所有 L 型中心部分粘结在一起作为中柱。为了提高其整体刚度以抵抗扭转载荷，并避免脱粘失效，中柱由具有足够预应力的绝缘圆筒包裹。此外，由于希尔思线圈是缠绕在绝缘圆筒上的，因此绝缘圆筒需要有良好的尺寸精度。本文介绍了在绝缘筒开发过程中进行的技术研究结果，包括材料选择、参数计算、铺设试验和固化试验。
{"title":"Technical studies about the prestressed insulation cylinder of HL-3 tokamak's center-post","authors":"Yin Qiu, Ming Li, Xiaolong Liu, Qinghua Ren, Tao Lin, Jian Liu","doi":"10.1016/j.net.2024.08.001","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.001","url":null,"abstract":"The TF coil of HL-3 consists of 20 D-shaped demountable bundles, and all their L-shaped center sections are bonded together as the center-post. To increase its overall rigidity to resist torsional loads and avoid de-bonding failure, the center-post is wrapped by an insulation cylinder with enough prestress. Additionally, the insulation cylinder needs to be of good dimensional precision because the CS coil is winded around it. This paper describes the results of the technical studies carried out during the development of the insulation cylinder, including material selection, parameter calculation, enlacing tests and curing tests.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207274","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Deep geologic disposal has been widely accepted as a strategy for long-term disposal of the high-level radioactive waste. It is principal to obtain the migration parameters of radionuclides in natural barrier, such as granite, of a high-level radioactive waste repository for safety assessment of the repository. To quickly determine the diffusion and sorption properties of nuclides in intact granite, two tracers, I and ReO, were tested with a modified electromigration device, by imposing a constant voltage over an intact Beishan granitic rock sample. The breakthrough curves of I and ReO were obtained under condition of five different voltages. To interpret the electromigration experimental results with more confidence, an advection-dispersion model based on first-order adsorption kinetics was developed in this study. Data analysis of the breakthrough curves by this model suggest that the effective diffusion coefficients of I and ReO in intact Beishan granodiorite rock are (6.81 ± 0.53) × 10 m/s and (6.45 ± 0.07) × 10 m/s, respectively. While the distribution coefficient of the two ions are (9.06 ± 1.13) × 10 m/kg and (9.81 ± 0.13) × 10 m/kg, respectively. This indicates that I and ReO hardly adsorb in Beishan granodiorite rock.
深部地质处置作为一种长期处置高放射性废物的策略已被广泛接受。获取放射性核素在高放射性废物处置库天然屏障（如花岗岩）中的迁移参数是对处置库进行安全评估的关键。为了快速确定核素在完整花岗岩中的扩散和吸附特性，利用改进的电迁移装置，在完整的北山花岗岩样本上施加恒定电压，对 I 和 ReO 两种示踪剂进行了测试。在五种不同电压条件下获得了 I 和 ReO 的突破曲线。为了更有把握地解释电迁移实验结果，本研究建立了一个基于一阶吸附动力学的平流-分散模型。根据该模型对突破曲线进行的数据分析表明，I 和 ReO 在完整北山花岗闪长岩中的有效扩散系数分别为 (6.81 ± 0.53) × 10 m/s 和 (6.45 ± 0.07) × 10 m/s。而这两种离子的分布系数分别为（9.06 ± 1.13）×10 m/kg和（9.81 ± 0.13）×10 m/kg。这表明 I 和 ReO 在北山花岗闪长岩中几乎不吸附。
{"title":"Rapid determination of the migration parameters of nuclides in intact granite rock under the action of electric field","authors":"Xinyu Wang, Xiaojie Li, Yongmei Li, Longcheng Liu, Shuo Meng, Chunguang Li, Zhenzhong Liu, Xiaodong Li, Kaixuan Tan","doi":"10.1016/j.net.2024.07.062","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.062","url":null,"abstract":"Deep geologic disposal has been widely accepted as a strategy for long-term disposal of the high-level radioactive waste. It is principal to obtain the migration parameters of radionuclides in natural barrier, such as granite, of a high-level radioactive waste repository for safety assessment of the repository. To quickly determine the diffusion and sorption properties of nuclides in intact granite, two tracers, I and ReO, were tested with a modified electromigration device, by imposing a constant voltage over an intact Beishan granitic rock sample. The breakthrough curves of I and ReO were obtained under condition of five different voltages. To interpret the electromigration experimental results with more confidence, an advection-dispersion model based on first-order adsorption kinetics was developed in this study. Data analysis of the breakthrough curves by this model suggest that the effective diffusion coefficients of I and ReO in intact Beishan granodiorite rock are (6.81 ± 0.53) × 10 m/s and (6.45 ± 0.07) × 10 m/s, respectively. While the distribution coefficient of the two ions are (9.06 ± 1.13) × 10 m/kg and (9.81 ± 0.13) × 10 m/kg, respectively. This indicates that I and ReO hardly adsorb in Beishan granodiorite rock.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141940108","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-02DOI: 10.1016/j.net.2024.07.060
Woo-Min Cho, Han-Sang Woo, Yoon-Suk Chang
This study is to examine load-carrying capacity of a prestressed concrete containment vessel under the structural failure mode test non-linear finite element (FE) analyses. Firstly, suitability of three candidate structural failure criteria was evaluated at ambient temperature condition, of which results showed that the maximum principal strain-based one predicts ultimate pressure capacity (UPC) most closely with the test data. Effect of increasing temperature corresponding to a postulated severe accident-induced condition was investigated and the UPC exhibited reduction ratios of 1.19–1.49% at the peak temperature of 200 °C approximately depending on each failure criterion. Finally, parametric FE analyses at 95% confidence level were performed to quantify effect of material property uncertainties. Overall, the impact of altered material properties of concrete and rebar was higher than that of tendon prestress, and the increase of UPC in upper bound cases exceeded the decrease of UPC in lower bound cases.
本研究旨在研究预应力混凝土安全壳在结构失效模式试验非线性有限元（FE）分析下的承载能力。首先，在常温条件下评估了三种候选结构失效标准的适用性，结果表明，基于最大主应变的结构失效标准预测的极限承压能力（UPC）与试验数据最为接近。研究了温度升高对假定的严重事故诱发条件的影响，根据不同的失效标准，在峰值温度 200 °C 时，UPC 的降低率约为 1.19-1.49%。最后，进行了置信度为 95% 的参数 FE 分析，以量化材料特性不确定性的影响。总体而言，混凝土和钢筋的材料特性改变的影响大于肌腱预应力的影响，在上限情况下 UPC 的增加超过了下限情况下 UPC 的减少。
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