In this work, the thermal characteristics and steady-state temperatures (SST) of CPU and FPGA of electronic system in nuclear power plant are explored. Finite element analysis is performed to simulate the test process. Furthermore, three machine learning algorithms are used to predict chips temperatures at different operating conditions. It is found that when the ambient temperature is 20 °C and all the fans are power-off, the SST of the CPU and FPGA reaches 75 °C and 72 °C, respectively. While when the fans are power-on, the SST of the CPU and FPGA drops to 37.5 °C and 33 °C. When the ambient temperature increases to 55 °C and all the fans are power-on, the SST of the CPU and FPGA is 72.3 °C and 68.2 °C, respectively. The finite element model is verified and used to generate test data. Three machine learning models are verified by predicting the SST of CPU and FPGA under different operating conditions. It is found that M-SVR has better prediction ability than DT and ANN. The findings can be used for chip reliability evaluation of other electronic system devices, and provide a new method for predicting the possible steady-state temperature of chips under different service conditions.
{"title":"Investigation on the thermal characteristics of electronic system and prediction of chip temperature by machine learning","authors":"Fanyu Wang, Dongwei Wang, Qiang Deng, Hao Yan, Qi Chen, Yang Zhao","doi":"10.1016/j.net.2024.08.028","DOIUrl":"10.1016/j.net.2024.08.028","url":null,"abstract":"<div><div>In this work, the thermal characteristics and steady-state temperatures (SST) of CPU and FPGA of electronic system in nuclear power plant are explored. Finite element analysis is performed to simulate the test process. Furthermore, three machine learning algorithms are used to predict chips temperatures at different operating conditions. It is found that when the ambient temperature is 20 °C and all the fans are power-off, the SST of the CPU and FPGA reaches 75 °C and 72 °C, respectively. While when the fans are power-on, the SST of the CPU and FPGA drops to 37.5 °C and 33 °C. When the ambient temperature increases to 55 °C and all the fans are power-on, the SST of the CPU and FPGA is 72.3 °C and 68.2 °C, respectively. The finite element model is verified and used to generate test data. Three machine learning models are verified by predicting the SST of CPU and FPGA under different operating conditions. It is found that M-SVR has better prediction ability than DT and ANN. The findings can be used for chip reliability evaluation of other electronic system devices, and provide a new method for predicting the possible steady-state temperature of chips under different service conditions.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103159"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207251","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 10.1016/j.net.2024.08.015
M.I. Sayyed , K.A. Mahmoud , Taha A. Hanafy
This work presents a study of the effect of replacing lead dioxide and gadolinium (III) oxide with boron trioxide on the physical, mechanical, and radiation shielding properties for the B2O3-Na2O-ZnO-PbO2-Gd2O3 glass systems. The Archimedes method confirms that the increase in the PbO2+Gd2O3 concentration within the fabricated glass system in the range from 16 to 22 mol.% increases the fabricated glass samples' density from 4.052 to 4.408 g/cm3. Additionally, the Makishima-Mackenzie model was utilized to investigate the influence of PbO2+Gd2O3 on the mechanical properties of the investigated glass samples. The increase in the substituting of PbO2+Gd2O3 decreases the fabricated glass samples' mechanical properties and micro-hardness. Furthermore, the Monte Carlo simulation method was applied for the estimation of the impact of PbO2+Gd2O3 concentration on the fabricated samples' radiation shielding parameters. The increase in the concentration of PbO2+Gd2O3 with range of (16, 18, 20 and 22) leads to increase the linear attenuation coefficient (LAC) to 8.014–11.517 cm−1 at 0.06 MeV, 0.381–0.423 cm−1 at 0.6 MeV, 0.133–0.149 cm−1 at 5 MeV, and 0.132–0.154 cm−1 at 15 MeV with the same order, respectively. Therefore, the introduction of PbO2+Gd2O3 concentration enhances the fabricated glass samples' radiation shielding properties to be suitable for γ-ray shielding applications.
{"title":"A new protective glass material against gamma ray: Thorough analysis to determine the impact of adding gadolinium (III) oxide","authors":"M.I. Sayyed , K.A. Mahmoud , Taha A. Hanafy","doi":"10.1016/j.net.2024.08.015","DOIUrl":"10.1016/j.net.2024.08.015","url":null,"abstract":"<div><div>This work presents a study of the effect of replacing lead dioxide and gadolinium (III) oxide with boron trioxide on the physical, mechanical, and radiation shielding properties for the B<sub>2</sub>O<sub>3</sub>-Na<sub>2</sub>O-ZnO-PbO<sub>2</sub>-Gd<sub>2</sub>O<sub>3</sub> glass systems. The Archimedes method confirms that the increase in the PbO<sub>2</sub>+Gd<sub>2</sub>O<sub>3</sub> concentration within the fabricated glass system in the range from 16 to 22 mol.% increases the fabricated glass samples' density from 4.052 to 4.408 g/cm<sup>3</sup>. Additionally, the Makishima-Mackenzie model was utilized to investigate the influence of PbO<sub>2</sub>+Gd<sub>2</sub>O<sub>3</sub> on the mechanical properties of the investigated glass samples. The increase in the substituting of PbO<sub>2</sub>+Gd<sub>2</sub>O<sub>3</sub> decreases the fabricated glass samples' mechanical properties and micro-hardness. Furthermore, the Monte Carlo simulation method was applied for the estimation of the impact of PbO<sub>2</sub>+Gd<sub>2</sub>O<sub>3</sub> concentration on the fabricated samples' radiation shielding parameters. The increase in the concentration of PbO<sub>2</sub>+Gd<sub>2</sub>O<sub>3</sub> with range of (16, 18, 20 and 22) leads to increase the linear attenuation coefficient (LAC) to 8.014–11.517 cm<sup>−1</sup> at 0.06 MeV, 0.381–0.423 cm<sup>−1</sup> at 0.6 MeV, 0.133–0.149 cm<sup>−1</sup> at 5 MeV, and 0.132–0.154 cm<sup>−1</sup> at 15 MeV with the same order, respectively. Therefore, the introduction of PbO<sub>2</sub>+Gd<sub>2</sub>O<sub>3</sub> concentration enhances the fabricated glass samples' radiation shielding properties to be suitable for γ-ray shielding applications.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103146"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207266","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 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":"10.1016/j.net.2024.08.007","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103138"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","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":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 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":"10.1016/j.net.2024.08.010","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103141"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","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":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 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 via 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。
{"title":"Investigation on structural failure criteria and material property uncertainties of prestressed concrete containment structure","authors":"Woo-Min Cho , Han-Sang Woo , Yoon-Suk Chang","doi":"10.1016/j.net.2024.07.060","DOIUrl":"10.1016/j.net.2024.07.060","url":null,"abstract":"<div><div>This study is to examine load-carrying capacity of a prestressed concrete containment vessel under the structural failure mode test <em>via</em> 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.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103129"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141969113","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 10.1016/j.net.2024.07.059
Eunjung Cho , Juyong Lee
This study provides a comprehensive analysis of the social acceptance of the Small Modular Reactor (SMR) in South Korea. The growing interest in SMRs as a potential solution to the challenges of climate change and energy security highlights the need for continued research and development in this area. The research was conducted using a contingent valuation method, which involved surveying a representative sample of the South Korean population. Out of a total of 1032 respondents, 78 % were willing to pay for SMR development. The mean WTP was estimated to be USD 4.28 per year per household and scaled up to the national level in South Korea, it was analysed to have a total economic and social benefit of USD 0.49 billion. The more serious the respondents perceived the current and future problem of climate change to be, the more likely they were to be willing to pay an offered bid amount. This study suggests that the need to address climate change and the need to raise awareness of SMR as one of the low-carbon technologies at the government level is required to generate the appropriate level of social acceptance to promote SMR development.
{"title":"Social acceptance of small modular reactor (SMR): Evidence from a contingent valuation study in South Korea","authors":"Eunjung Cho , Juyong Lee","doi":"10.1016/j.net.2024.07.059","DOIUrl":"10.1016/j.net.2024.07.059","url":null,"abstract":"<div><div>This study provides a comprehensive analysis of the social acceptance of the Small Modular Reactor (SMR) in South Korea. The growing interest in SMRs as a potential solution to the challenges of climate change and energy security highlights the need for continued research and development in this area. The research was conducted using a contingent valuation method, which involved surveying a representative sample of the South Korean population. Out of a total of 1032 respondents, 78 % were willing to pay for SMR development. The mean WTP was estimated to be USD 4.28 per year per household and scaled up to the national level in South Korea, it was analysed to have a total economic and social benefit of USD 0.49 billion. The more serious the respondents perceived the current and future problem of climate change to be, the more likely they were to be willing to pay an offered bid amount. This study suggests that the need to address climate change and the need to raise awareness of SMR as one of the low-carbon technologies at the government level is required to generate the appropriate level of social acceptance to promote SMR development.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103128"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141940107","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 10.1016/j.net.2024.08.029
Xin Wang , Ze Xu , Yulan Liu , Biao Wang
The elastoplastic mechanical behavior of Zircaloy-4 (Zr-4) cladding, coated with chromium (Cr) or FeCrAl on its surface, is explored under the coupled effects of multi-field coupling. Utilizing the Finite Element Software ABAQUS, simulations are conducted to calculate the evolution of stress and strain over two complete fuel cycles. Comparisons are drawn between the coated and uncoated Zircaloy-4 cladding materials. The results indicate that the application of surface coatings significantly mitigates stress levels in the cladding during the first fuel cycle. During the second fuel cycle, all three types of cladding exhibit relatively minor plastic strain, which is attributed to the unloading and reloading process between cycles. Notably, the plastic zone propagates from the interior to the exterior of the cladding. When compared to traditional Zircaloy-4 cladding, the coated cladding exhibits improved elastoplastic mechanical behavior. The operational mechanism of the coating for different stresses in cylindrical coordinates and its response to unloading and reloading cycles are also investigated. Specifically, the coated claddings exhibit an evident delay in reaching full plasticity compared to uncoated claddings. Furthermore, FeCrAl coating material initially shows good performance, and it needs to be verified in more aspects in the future. Results and Conclusions in this paper can provide reference and guidance for future experiments.
{"title":"Elastoplastic mechanical behavior analysis of surface-coated Zircaloy-4 cladding under multi-field coupling","authors":"Xin Wang , Ze Xu , Yulan Liu , Biao Wang","doi":"10.1016/j.net.2024.08.029","DOIUrl":"10.1016/j.net.2024.08.029","url":null,"abstract":"<div><div>The elastoplastic mechanical behavior of Zircaloy-4 (Zr-4) cladding, coated with chromium (Cr) or FeCrAl on its surface, is explored under the coupled effects of multi-field coupling. Utilizing the Finite Element Software ABAQUS, simulations are conducted to calculate the evolution of stress and strain over two complete fuel cycles. Comparisons are drawn between the coated and uncoated Zircaloy-4 cladding materials. The results indicate that the application of surface coatings significantly mitigates stress levels in the cladding during the first fuel cycle. During the second fuel cycle, all three types of cladding exhibit relatively minor plastic strain, which is attributed to the unloading and reloading process between cycles. Notably, the plastic zone propagates from the interior to the exterior of the cladding. When compared to traditional Zircaloy-4 cladding, the coated cladding exhibits improved elastoplastic mechanical behavior. The operational mechanism of the coating for different stresses in cylindrical coordinates and its response to unloading and reloading cycles are also investigated. Specifically, the coated claddings exhibit an evident delay in reaching full plasticity compared to uncoated claddings. Furthermore, FeCrAl coating material initially shows good performance, and it needs to be verified in more aspects in the future. Results and Conclusions in this paper can provide reference and guidance for future experiments.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103160"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207252","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 10.1016/j.net.2024.08.036
Xinbo Li , Jinxin Gong
This paper assesses the aging effects on the leak-tightness of containment under overpressure conditions. A global containment model and the detailed sub-models for the three main penetration regions in the containment are established. The main aging forms and mechanisms of containment are clarified, and corresponding simulation schemes are provided. Through numerical simulation, the impacts of different aging forms on the failure pressure of containment are discussed from a deterministic perspective. Finally, the fragility and functional failure probability of containment under different aging conditions are evaluated. When 60 years of concrete degradation and prestress loss are considered, the pressure capacities of equipment hatch, personnel airlock, and pipe penetration are reduced by approximately 5 %. Upon further considering steel liner corrosion, when the corrosion degree reaches 30 %, the pressure capacities of these regions are reduced by 24.37 %, 25.52 %, and 22.83 %, respectively. Within the scope of this study, the impact of steel liner corrosion on the risk of containment leakage is the most pronounced, whereas the impact of concrete degradation is minimal. If steel liner corrosion occurs simultaneously in three penetration regions, the whole containment will fail to meet the probabilistic performance goal when the corrosion degree reaches 30 %.
{"title":"Assessment of leak-tightness for nuclear reactor containment under overpressure conditions considering aging effects","authors":"Xinbo Li , Jinxin Gong","doi":"10.1016/j.net.2024.08.036","DOIUrl":"10.1016/j.net.2024.08.036","url":null,"abstract":"<div><div>This paper assesses the aging effects on the leak-tightness of containment under overpressure conditions. A global containment model and the detailed sub-models for the three main penetration regions in the containment are established. The main aging forms and mechanisms of containment are clarified, and corresponding simulation schemes are provided. Through numerical simulation, the impacts of different aging forms on the failure pressure of containment are discussed from a deterministic perspective. Finally, the fragility and functional failure probability of containment under different aging conditions are evaluated. When 60 years of concrete degradation and prestress loss are considered, the pressure capacities of equipment hatch, personnel airlock, and pipe penetration are reduced by approximately 5 %. Upon further considering steel liner corrosion, when the corrosion degree reaches 30 %, the pressure capacities of these regions are reduced by 24.37 %, 25.52 %, and 22.83 %, respectively. Within the scope of this study, the impact of steel liner corrosion on the risk of containment leakage is the most pronounced, whereas the impact of concrete degradation is minimal. If steel liner corrosion occurs simultaneously in three penetration regions, the whole containment will fail to meet the probabilistic performance goal when the corrosion degree reaches 30 %.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103167"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207244","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The urgent requirement for sustainable and dependable energy sources has stimulated an increased fascination with precisely forecasting nuclear energy generation. This work utilizes sophisticated regression modeling approaches, namely XGBoost, to predict nuclear energy generation by leveraging economic indices such as Gross Domestic Product (GDP). Each model's prediction accuracy has been evaluated by examining historical data on nuclear energy output and GDP from various locations. Here, measures such as mean squared error (MSE) and coefficient of determination (R2) to analyze their effectiveness have been used. The results of this study demonstrate that the XGBoost model outperforms standard regression approaches, showing greater R2 values and lower MSE scores. Furthermore, the consequences of these discoveries for the development of energy policy offer possible directions for future study in energy forecasting. This study provides useful insights for energy planners and policymakers, enabling a more profound comprehension of the complex relationship between economic indicators and nuclear energy generation.
{"title":"Advancing nuclear energy forecasting: Exploring regression modeling techniques for improved accuracy","authors":"Anjali Nighoskar , Preeti Chaurasia , Nagendra Singh","doi":"10.1016/j.net.2024.08.013","DOIUrl":"10.1016/j.net.2024.08.013","url":null,"abstract":"<div><div>The urgent requirement for sustainable and dependable energy sources has stimulated an increased fascination with precisely forecasting nuclear energy generation. This work utilizes sophisticated regression modeling approaches, namely XGBoost, to predict nuclear energy generation by leveraging economic indices such as Gross Domestic Product (GDP). Each model's prediction accuracy has been evaluated by examining historical data on nuclear energy output and GDP from various locations. Here, measures such as mean squared error (MSE) and coefficient of determination (R2) to analyze their effectiveness have been used. The results of this study demonstrate that the XGBoost model outperforms standard regression approaches, showing greater R2 values and lower MSE scores. Furthermore, the consequences of these discoveries for the development of energy policy offer possible directions for future study in energy forecasting. This study provides useful insights for energy planners and policymakers, enabling a more profound comprehension of the complex relationship between economic indicators and nuclear energy generation.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":"Article 103144"},"PeriodicalIF":2.6,"publicationDate":"2025-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207248","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-01-01DOI: 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.
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