Pub Date : 2024-08-15DOI: 10.1016/j.net.2024.08.023
Haoxiang Li, Wei Zheng, Bin Du, Huang Zhang, Huaqiang Yin, Xuedong He, Tao Ma, Xingtuan Yang
Air ingress accident is one of the typical accident conditions in Very High Temperature Reactors (VHTRs). This work investigates the oxidation kinetics, corrosion behavior and mechanism of Inconel 617 alloy in different oxygen concentration atmospheres under air ingress accident. The impact of O concentration and oxidation time of the alloy corrosion is investigated. A gas chromatograph was used to measure the impurity content in real time during the helium experiments. After the experiments, the alloys were characterized by electronic balance, were analyzed by Scanning Electron Microscopy (SEM), Energy-Dispersive X-ray Spectroscopy (EDS), X-ray diffraction (XRD) and carbon and sulfur analyzer. The results show that: the Inconel 617 alloy undergoes similar oxidation behavior and the degree of oxidation is very close in three groups of atmospheres with large differences in oxygen content; the alloy should show two oxidation mechanisms, linear oxidation and parabolic oxidation, during the oxidation process; the parabolic rate constant and of the alloy is a constant value and does not vary with large changes in oxygen concentration, but when the experimental temperature changes, the oxidation rate constants of the alloy change, and the lowering of the temperature leads to the lowering of the oxidation rate constants; When experimental temperature is at 950 °C, the alloy continues to undergo a “microclimatic reaction” in the atmosphere of He-ppmO, and the microclimatic reaction disappears when the experimental temperature is lowered to 750 °C; In He-ppmO environment, gas chromatograph can be used instead of thermogravimetric analyzer for real-time monitoring.
{"title":"Corrosion studies of Inconel 617 in high temperature air and He-ppmO2 atmospheres","authors":"Haoxiang Li, Wei Zheng, Bin Du, Huang Zhang, Huaqiang Yin, Xuedong He, Tao Ma, Xingtuan Yang","doi":"10.1016/j.net.2024.08.023","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.023","url":null,"abstract":"Air ingress accident is one of the typical accident conditions in Very High Temperature Reactors (VHTRs). This work investigates the oxidation kinetics, corrosion behavior and mechanism of Inconel 617 alloy in different oxygen concentration atmospheres under air ingress accident. The impact of O concentration and oxidation time of the alloy corrosion is investigated. A gas chromatograph was used to measure the impurity content in real time during the helium experiments. After the experiments, the alloys were characterized by electronic balance, were analyzed by Scanning Electron Microscopy (SEM), Energy-Dispersive X-ray Spectroscopy (EDS), X-ray diffraction (XRD) and carbon and sulfur analyzer. The results show that: the Inconel 617 alloy undergoes similar oxidation behavior and the degree of oxidation is very close in three groups of atmospheres with large differences in oxygen content; the alloy should show two oxidation mechanisms, linear oxidation and parabolic oxidation, during the oxidation process; the parabolic rate constant and of the alloy is a constant value and does not vary with large changes in oxygen concentration, but when the experimental temperature changes, the oxidation rate constants of the alloy change, and the lowering of the temperature leads to the lowering of the oxidation rate constants; When experimental temperature is at 950 °C, the alloy continues to undergo a “microclimatic reaction” in the atmosphere of He-ppmO, and the microclimatic reaction disappears when the experimental temperature is lowered to 750 °C; In He-ppmO environment, gas chromatograph can be used instead of thermogravimetric analyzer for real-time monitoring.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"25 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207253","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}
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":"https://doi.org/10.1016/j.net.2024.08.028","url":null,"abstract":"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.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"47 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-15","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":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-14DOI: 10.1016/j.net.2024.08.025
Danwoo Ko, Seunguk Cheon, Jiyoung Kim, Seungmin Lee, Seung Min Woo
There is an argument that the Comprehensive Safeguards Agreement (CSA) should be applied to the milling process, which is the first stage of nuclear fuel cycle, to reduce the possibility of nuclear proliferation. Therefore, this study aims to propose new and conceptual safeguards applicable to milling facilities and design Nuclear Material Accountancy for its application. Mill tailings reach a secular equilibrium state. While approaching this equilibrium state, the ratio of each isotope changes. First, the Bateman equation was used to analyze this phenomenon and evaluate the feasibility of utilizing the specific isotope ratio to track undeclared nuclear activities. Second, the gamma spectrum analysis of mill tailings was conducted using the Monte Carlo N-Particle Code to validate undeclared nuclear activities. This study shows that the ratio of U-235/Th-234 and U-235/Pa-234m can estimate the production time within a year. Furthermore, gamma spectrum analysis of mill tailings revealed visible differences in the low-energy region due to the decay of Th-234. Finally, a conceptual Material Balance Area, Key Measurement Points, and Material Balance Period for milling facilities were designed to apply CSA. It is anticipated that applying CSA to milling facilities could enhance not only the nuclear fuel cycles but also nuclear non-proliferation system.
{"title":"Conceptual safeguards method proposal for milling facilities based on nuclear isotopic ratios in uranium mill tailings","authors":"Danwoo Ko, Seunguk Cheon, Jiyoung Kim, Seungmin Lee, Seung Min Woo","doi":"10.1016/j.net.2024.08.025","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.025","url":null,"abstract":"There is an argument that the Comprehensive Safeguards Agreement (CSA) should be applied to the milling process, which is the first stage of nuclear fuel cycle, to reduce the possibility of nuclear proliferation. Therefore, this study aims to propose new and conceptual safeguards applicable to milling facilities and design Nuclear Material Accountancy for its application. Mill tailings reach a secular equilibrium state. While approaching this equilibrium state, the ratio of each isotope changes. First, the Bateman equation was used to analyze this phenomenon and evaluate the feasibility of utilizing the specific isotope ratio to track undeclared nuclear activities. Second, the gamma spectrum analysis of mill tailings was conducted using the Monte Carlo N-Particle Code to validate undeclared nuclear activities. This study shows that the ratio of U-235/Th-234 and U-235/Pa-234m can estimate the production time within a year. Furthermore, gamma spectrum analysis of mill tailings revealed visible differences in the low-energy region due to the decay of Th-234. Finally, a conceptual Material Balance Area, Key Measurement Points, and Material Balance Period for milling facilities were designed to apply CSA. It is anticipated that applying CSA to milling facilities could enhance not only the nuclear fuel cycles but also nuclear non-proliferation system.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"13 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207254","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 : 2024-08-13DOI: 10.1016/j.net.2024.08.020
Hyeongjin Byeon, Ugyu Jeong, Jaeyeong Park
Homogeneity is an important factor for ensuring the structural stability of solidified radioactive waste, and the most effective approach for assessing its homogeneity is by performing compressive strength measurements using the minimum amount of coring specimens. The efficiency of detecting inhomogeneous waste is affected by the coring position and number of coring positions. However, no guidelines exist for coring solidified waste for compressive-strength tests. Therefore, this study compared uniform, random, and quasi-Monte Carlo sampling methods to determine the most effective core position. Further, the effects of different sampling amounts on the detection rate of inhomogeneous solidified waste were observed, and the detection rate of the inhomogeneous waste was obtained by modeling the coring procedure of solidified radioactive waste using MATLAB. Thus, a sampling method and a method for increasing the specimen amount, both of which can efficiently detect inhomogeneous waste during compressive strength tests, were presented in this paper. The results of this study can be applied as background data for developing homogeneity assessment guidelines for solidified radioactive waste.
{"title":"Statistical sampling method to verify the homogeneity of full-scale cement-solidified radioactive waste","authors":"Hyeongjin Byeon, Ugyu Jeong, Jaeyeong Park","doi":"10.1016/j.net.2024.08.020","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.020","url":null,"abstract":"Homogeneity is an important factor for ensuring the structural stability of solidified radioactive waste, and the most effective approach for assessing its homogeneity is by performing compressive strength measurements using the minimum amount of coring specimens. The efficiency of detecting inhomogeneous waste is affected by the coring position and number of coring positions. However, no guidelines exist for coring solidified waste for compressive-strength tests. Therefore, this study compared uniform, random, and quasi-Monte Carlo sampling methods to determine the most effective core position. Further, the effects of different sampling amounts on the detection rate of inhomogeneous solidified waste were observed, and the detection rate of the inhomogeneous waste was obtained by modeling the coring procedure of solidified radioactive waste using MATLAB. Thus, a sampling method and a method for increasing the specimen amount, both of which can efficiently detect inhomogeneous waste during compressive strength tests, were presented in this paper. The results of this study can be applied as background data for developing homogeneity assessment guidelines for solidified radioactive waste.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"5 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207255","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 : 2024-08-13DOI: 10.1016/j.net.2024.08.021
Seok-Min Hong, Chang Hwa Lee, Jae-Hwan Yang, Namcheol Kim
The continuous utilization of nuclear energy has led to the accumulation of spent nuclear fuel (SNF) containing uranium, transuranium, and fission products (FPs). Reprocessing and pyrochemical methods have shown the potential for SNF reuse, thereby reducing its environmental impact. Voloxidation, a pivotal step in SNF recycling, involves thermal treatment in an oxidizing atmosphere to enhance the reactivity. During voloxidation, Se-79, which is a FP with a long half-life, is released as SeO under oxidizing conditions, necessitating technologies to capture it. CaO pellets (CPs) were prepared to capture gaseous SeO. The effects of operating conditions on SeO capture performance were investigated. The CP reacts strongly with SeO to form CaSeO, exhibiting a high adsorption capacity of 17.5 mol kg and >99 % efficiency at 700 °C. The mechanical strength and thermal stability assessments indicate suitability for practical applications. Depending on the flow rate, the CP required for SeO capture was estimated when processing 1 t of SNF, thereby contributing to the design of effective SeO capture systems for safe and sustainable nuclear waste management.
核能的持续利用导致了含有铀、超铀和裂变产物（FPs）的乏核燃料（SNF）的积累。后处理和热化学方法显示了乏核燃料再利用的潜力，从而减少了其对环境的影响。氧化是 SNF 循环利用的关键步骤，包括在氧化气氛中进行热处理，以提高反应活性。在氧化过程中，Se-79 是一种半衰期较长的 FP，在氧化条件下会以 SeO 的形式释放出来，因此需要采用技术来捕获它。为捕获气态 SeO，制备了 CaO 颗粒（CPs）。研究了操作条件对氧化硒捕获性能的影响。CP 与 SeO 发生强烈反应，生成 CaSeO，在 700 °C 时的吸附容量高达 17.5 mol kg，吸附效率大于 99%。机械强度和热稳定性评估表明，这种材料适合实际应用。根据流速的不同，估算了处理 1 吨 SNF 时捕获 SeO 所需的氯化石蜡，从而有助于设计有效的 SeO 捕获系统，实现安全、可持续的核废料管理。
{"title":"Effective capture of gaseous Se during spent nuclear fuel recycling using calcium oxide pellets: Optimization, performance, and practical implications","authors":"Seok-Min Hong, Chang Hwa Lee, Jae-Hwan Yang, Namcheol Kim","doi":"10.1016/j.net.2024.08.021","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.021","url":null,"abstract":"The continuous utilization of nuclear energy has led to the accumulation of spent nuclear fuel (SNF) containing uranium, transuranium, and fission products (FPs). Reprocessing and pyrochemical methods have shown the potential for SNF reuse, thereby reducing its environmental impact. Voloxidation, a pivotal step in SNF recycling, involves thermal treatment in an oxidizing atmosphere to enhance the reactivity. During voloxidation, Se-79, which is a FP with a long half-life, is released as SeO under oxidizing conditions, necessitating technologies to capture it. CaO pellets (CPs) were prepared to capture gaseous SeO. The effects of operating conditions on SeO capture performance were investigated. The CP reacts strongly with SeO to form CaSeO, exhibiting a high adsorption capacity of 17.5 mol kg and >99 % efficiency at 700 °C. The mechanical strength and thermal stability assessments indicate suitability for practical applications. Depending on the flow rate, the CP required for SeO capture was estimated when processing 1 t of SNF, thereby contributing to the design of effective SeO capture systems for safe and sustainable nuclear waste management.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"5 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207311","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 : 2024-08-13DOI: 10.1016/j.net.2024.08.022
A. Abdelghafar Galahom, Nassar Alnassar, Amr Ibrahim
This work investigates the possibility of getting rid of the accumulated plutonium around the world by mixing it with fertile materials and using it as a nuclear fuel for CANDU-6. Using plutonium isotopes with thorium as a nuclear fuel for CANDU-6 is a means of preventing the production of other Pu isotopes. MCNPX has been used to design a three dimensional model of the CANDU-6 bundle. Three fuel types including (U, rgPu)O, (Th, rgPu)O and (Th, wgPu)O have been examined as nuclear fuel in the designed model and their results were compared with UO as a standard fuel. The fuel burnup parameters such as k, fissile inventory ratio, plutonium concentration and minor actinides concentration have been analyzed for the suggested fuels. Some of the most related safety parameters such as effective delayed neutrons (β), moderator temperature coefficient, coolant temperature coefficient and Doppler constant have been studied. The excess thermal neutrons in the CANDU-6 maximized the benefit of using plutonium-based fuels, where a significant amount of plutonium has been burned during the fuel cycle. From the neutronic and safety point of view, thorium fuel mixed with reactor grade plutonium has proven to be the most promising candidate among the investigated fuels.
{"title":"Searching for incinerating the accumulated plutonium around the world by mixing it with thorium and using this mixture as a nuclear fuel in the CANDU-6","authors":"A. Abdelghafar Galahom, Nassar Alnassar, Amr Ibrahim","doi":"10.1016/j.net.2024.08.022","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.022","url":null,"abstract":"This work investigates the possibility of getting rid of the accumulated plutonium around the world by mixing it with fertile materials and using it as a nuclear fuel for CANDU-6. Using plutonium isotopes with thorium as a nuclear fuel for CANDU-6 is a means of preventing the production of other Pu isotopes. MCNPX has been used to design a three dimensional model of the CANDU-6 bundle. Three fuel types including (U, rgPu)O, (Th, rgPu)O and (Th, wgPu)O have been examined as nuclear fuel in the designed model and their results were compared with UO as a standard fuel. The fuel burnup parameters such as k, fissile inventory ratio, plutonium concentration and minor actinides concentration have been analyzed for the suggested fuels. Some of the most related safety parameters such as effective delayed neutrons (β), moderator temperature coefficient, coolant temperature coefficient and Doppler constant have been studied. The excess thermal neutrons in the CANDU-6 maximized the benefit of using plutonium-based fuels, where a significant amount of plutonium has been burned during the fuel cycle. From the neutronic and safety point of view, thorium fuel mixed with reactor grade plutonium has proven to be the most promising candidate among the investigated fuels.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"77 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207256","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 : 2024-08-13DOI: 10.1016/j.net.2024.08.018
Soonwoo Han, Jae-Seon Lee
In this study, we proposed in-vessel two-channel control rod positioner models for small nuclear reactors and carried out performance tests to evaluate applicability of the models. Embedding control rod positioners in reactors reduces the volume of small reactors and minimizes the penetration parts of reactor. In this study, two models having two physically separated solenoids were developed for measuring the position of the control rod by using the inductance change of a solenoid. First, the principle of measuring the position of a ferromagnetic rod by using a solenoid was formulated, and the change in solenoid inductance with the rod position was calculated through finite element analysis. Based on the proposed models, test products were manufactured and tested on three items to confirm the feasibility of the proposed solenoid-based two-channel control rod position indicator.
{"title":"Development and experimental validation of conceptual models for solenoid-based in-vessel two-channel control rod position indicator","authors":"Soonwoo Han, Jae-Seon Lee","doi":"10.1016/j.net.2024.08.018","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.018","url":null,"abstract":"In this study, we proposed in-vessel two-channel control rod positioner models for small nuclear reactors and carried out performance tests to evaluate applicability of the models. Embedding control rod positioners in reactors reduces the volume of small reactors and minimizes the penetration parts of reactor. In this study, two models having two physically separated solenoids were developed for measuring the position of the control rod by using the inductance change of a solenoid. First, the principle of measuring the position of a ferromagnetic rod by using a solenoid was formulated, and the change in solenoid inductance with the rod position was calculated through finite element analysis. Based on the proposed models, test products were manufactured and tested on three items to confirm the feasibility of the proposed solenoid-based two-channel control rod position indicator.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"38 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142226267","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 : 2024-08-12DOI: 10.1016/j.net.2024.08.019
Jacob G. Fantidis, Athanasia Κ. Thomoglou, Maristella Ε. Voutetaki, Bassam Tayeh, George Nikolaou
The evaluation of non-destructive neutron radiography (NR) for examining the internal composition of various structural materials, has been the focus of extensive research. This manuscript uses non-destructive testing to generate three-dimensional radiographs of three different brick structural materials: glass block, magnesia-chrome, and lead to evaluate their capability to withstand fast neutrons and gammas emitted from a source. When neutrons with thermal or epithermal spectrum are required, the optimum combination for an accelerator was simulated using a 2.8 MeV proton beam on a lithium target. The presented facility tested both thermal and fast neutron radiography. This study examined various aperture diameters and collimator lengths. It found that implementing a special fast neutron filter significantly increased the thermal neutron content (TNC) with minimal impact on the thermal neutron flux. For fast neutron radiography, the study evaluated parameters such as geometric unsharpness, fast neutron flux, and the percentage of the uncollided fast neutron reaching the object. Both neutrons and photons from the source were used to inspect faults in a glass brick.
{"title":"Bricks non-destructive simulation testing method utilizing neutron radiography facility based on a 7Li(p,n)7Be reaction","authors":"Jacob G. Fantidis, Athanasia Κ. Thomoglou, Maristella Ε. Voutetaki, Bassam Tayeh, George Nikolaou","doi":"10.1016/j.net.2024.08.019","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.019","url":null,"abstract":"The evaluation of non-destructive neutron radiography (NR) for examining the internal composition of various structural materials, has been the focus of extensive research. This manuscript uses non-destructive testing to generate three-dimensional radiographs of three different brick structural materials: glass block, magnesia-chrome, and lead to evaluate their capability to withstand fast neutrons and gammas emitted from a source. When neutrons with thermal or epithermal spectrum are required, the optimum combination for an accelerator was simulated using a 2.8 MeV proton beam on a lithium target. The presented facility tested both thermal and fast neutron radiography. This study examined various aperture diameters and collimator lengths. It found that implementing a special fast neutron filter significantly increased the thermal neutron content (TNC) with minimal impact on the thermal neutron flux. For fast neutron radiography, the study evaluated parameters such as geometric unsharpness, fast neutron flux, and the percentage of the uncollided fast neutron reaching the object. Both neutrons and photons from the source were used to inspect faults in a glass brick.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"9 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207262","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}
The fluidelastic instability (FEI) in heat exchanger tubes has been of widespread concern due to its tendency to cause damage to the tubes. Generally, FEI in the transverse direction of the tube occurs earlier than in the streamwise direction, and the intrinsic frequency of the tube as well as the way of tube distribution have a great influence. The mechanisms involved in inducing FEI need to be further investigated. We set up an air-water two-phase flow water tunnel test system and adopt a normal triangular arrangement plate with a pitch-to-diameter ratio of 1.41 to conduct experiments. It was ensured that FEI could occur in the experimental flow range, by varying the intrinsic frequency of the flexible tube. The fluidelastic instability phenomenon was investigated in a single flexible tube array and a cluster of seven flexible tubes (the central cluster). Comparative analysis was conducted between the results of the two array configurations. The experiments involved concurrent streamwise and transverse directions, and an analysis of the pressure exerted on the flexible tubes. Additionally, the “transitional” state observed in flexible tubes under strongly coupled vibration in the central cluster was explored. The findings indicated that a cluster of seven flexible tubes intensified the vibration coupling between bundles, leading to a more complex flow field around the tube bundle perimeter, consequently exacerbating tube vibration. Furthermore, under the “transitional” state, enhanced stability was manifested. Additionally, instability in the streamwise direction was primarily controlled by the stiffness mechanism; a single flexible tube did not exhibit fluidelastic instability in the streamwise direction, while the central cluster did. This paper recommends an instability constant (K) value of 3.4 for the Connors formula for a normal triangular tube bundle configuration with the pitch-to-diameter ratio of 1.41, thereby providing empirical and theoretical support for the vibration analysis of tube bundles.
{"title":"Study of fluidelastic instability in the streamwise and transverse directions through tube array under two-phase flow conditions using pressure effects","authors":"Sijiu Qi, Wei Tan, Ke Zhang, Yuancen Wang, Wenjing Lin, Peize Han, Guorui Zhu","doi":"10.1016/j.net.2024.08.017","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.017","url":null,"abstract":"The fluidelastic instability (FEI) in heat exchanger tubes has been of widespread concern due to its tendency to cause damage to the tubes. Generally, FEI in the transverse direction of the tube occurs earlier than in the streamwise direction, and the intrinsic frequency of the tube as well as the way of tube distribution have a great influence. The mechanisms involved in inducing FEI need to be further investigated. We set up an air-water two-phase flow water tunnel test system and adopt a normal triangular arrangement plate with a pitch-to-diameter ratio of 1.41 to conduct experiments. It was ensured that FEI could occur in the experimental flow range, by varying the intrinsic frequency of the flexible tube. The fluidelastic instability phenomenon was investigated in a single flexible tube array and a cluster of seven flexible tubes (the central cluster). Comparative analysis was conducted between the results of the two array configurations. The experiments involved concurrent streamwise and transverse directions, and an analysis of the pressure exerted on the flexible tubes. Additionally, the “transitional” state observed in flexible tubes under strongly coupled vibration in the central cluster was explored. The findings indicated that a cluster of seven flexible tubes intensified the vibration coupling between bundles, leading to a more complex flow field around the tube bundle perimeter, consequently exacerbating tube vibration. Furthermore, under the “transitional” state, enhanced stability was manifested. Additionally, instability in the streamwise direction was primarily controlled by the stiffness mechanism; a single flexible tube did not exhibit fluidelastic instability in the streamwise direction, while the central cluster did. This paper recommends an instability constant (K) value of 3.4 for the Connors formula for a normal triangular tube bundle configuration with the pitch-to-diameter ratio of 1.41, thereby providing empirical and theoretical support for the vibration analysis of tube bundles.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"10 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207263","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 : 2024-08-10DOI: 10.1016/j.net.2024.08.016
Yuanjie Fang, Minrui Fei, Hong Qian
Reactor power level control is an effective way to achieve load tracking of Pressurized Water Reactor (PWR) in a nuclear power station. A novel Nonlinear Generalized Predictive Control with Extended Kalman Filter (NGPC + EKF) is proposed to solve the problem that discrete predictive model mismatch in noisy environment. In this paper, an NGPC controller is developed to realize the reactor load tracking, and an EKF is used to estimate reactor states and suppress noise. Finally, the control methods of PID, MPC, NGPC and NGPC + EKF are compared by two simulation experiments, load tracking experiment and step response experiment. The load tracking experiment results show that NGPC + EKF method obtains better noise suppression ability and tracking effect. In the step response experiment, the proposed NGPC + EKF scheme is also proved to have better step response performance than others.
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