Cyclotron is a key scientific tool and indispensable research platform for conducting cutting-edge research in nuclear equipment development as well as innovative applications of nuclear technology. The shell component has a double-layer thick-walled structure with intricate ribs and high-density, full-penetration welded joints. The mitigation of welding deformation is of profound significance to the performance of the cyclotron. The thick plate joint has many welding layers which will be divided into several steps to complete the backing, filler, and cap welding. The equivalent transverse and longitudinal plastic strains of different layers were extracted by the thermo-elastic-plastic method. The welding deformation generated by each layer of weld can be predicted by using the equivalent plastic strain, and the total distortion can be accumulated layer by layer. Numerical simulation and experimental studies were conducted on the welding deformation of the double shell specimen, and the welding sequence and design of the welding fixture were discussed in detail. The digital photogrammetry system was used to monitor the deformation state of the welded parts in real-time. The measured deformation was compared with the simulation results. Ultimately, the deformation of the specimen is controlled at 2.64 mm. The proposed method can flexibly evaluate the impact of each welding layer on welding deformation for multiple welds, which can provide technical guidance for cyclotron engineering manufacturing.
{"title":"Numerical investigation of welding deformation diminution for double shell structure using the layered inherent strain","authors":"Cheng Li, Hua Zhai, Lianwei Zhu, Zhihong Liu, Jianguo Ma, Xiaofeng Zhu, Qiong Liu","doi":"10.1016/j.net.2024.09.011","DOIUrl":"https://doi.org/10.1016/j.net.2024.09.011","url":null,"abstract":"Cyclotron is a key scientific tool and indispensable research platform for conducting cutting-edge research in nuclear equipment development as well as innovative applications of nuclear technology. The shell component has a double-layer thick-walled structure with intricate ribs and high-density, full-penetration welded joints. The mitigation of welding deformation is of profound significance to the performance of the cyclotron. The thick plate joint has many welding layers which will be divided into several steps to complete the backing, filler, and cap welding. The equivalent transverse and longitudinal plastic strains of different layers were extracted by the thermo-elastic-plastic method. The welding deformation generated by each layer of weld can be predicted by using the equivalent plastic strain, and the total distortion can be accumulated layer by layer. Numerical simulation and experimental studies were conducted on the welding deformation of the double shell specimen, and the welding sequence and design of the welding fixture were discussed in detail. The digital photogrammetry system was used to monitor the deformation state of the welded parts in real-time. The measured deformation was compared with the simulation results. Ultimately, the deformation of the specimen is controlled at 2.64 mm. The proposed method can flexibly evaluate the impact of each welding layer on welding deformation for multiple welds, which can provide technical guidance for cyclotron engineering manufacturing.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207207","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 core physical behavior of reactors is essentially the result of multi-physical fields coupling feedback. High-fidelity neutronics/thermal-hydraulics (N/TH) analysis can simulate and predict nuclear reactor core phenomena realistically, providing advanced and reliable technical means during the design and safety analysis of nuclear reactor. In this work, an efficient and robustness coupling method using power density as the coupling parameter, Matrix-Free Newton Krylov (MFNK) method, is successfully developed and innovatively implemented in HNET for high-fidelity N/TH coupling simulation. To enhance the efficiency and stability, the multi-level generalized equivalence theory-based CMFD (ML-gCMFD) iterative acceleration method and ML-gCMFD coupling acceleration method are proposed. In addition, the nonlinear preconditioning and hybrid perturbation size formula are implemented to further improve the convergence. Finally, to evaluate the numerical accuracy, convergence, efficiency and stability of MFNK method, a series of representative problems, including a three-dimensional (3D) single fuel pin problem, VERA Benchmark Problem 6, and VERA Benchmark Problem 7, are analyzed by comparing with the current N/TH coupling methods. Numerical results indicate that MFNK method can obtain strong stability, high convergence performance, and relatively high computational efficiency while ensuring high accuracy. It demonstrates that MFNK method has significant performance advantages and potential for high-fidelity N/TH coupling simulation.
{"title":"An innovative and efficient implementation of matrix-free Newton krylov method for neutronics/thermal-hydraulics coupling simulation","authors":"Peijun Li, Chen Hao, Ning Xu, Yanling Zhu, Yizhen Wang, Zhigang Zhang","doi":"10.1016/j.net.2024.09.012","DOIUrl":"https://doi.org/10.1016/j.net.2024.09.012","url":null,"abstract":"The core physical behavior of reactors is essentially the result of multi-physical fields coupling feedback. High-fidelity neutronics/thermal-hydraulics (N/TH) analysis can simulate and predict nuclear reactor core phenomena realistically, providing advanced and reliable technical means during the design and safety analysis of nuclear reactor. In this work, an efficient and robustness coupling method using power density as the coupling parameter, Matrix-Free Newton Krylov (MFNK) method, is successfully developed and innovatively implemented in HNET for high-fidelity N/TH coupling simulation. To enhance the efficiency and stability, the multi-level generalized equivalence theory-based CMFD (ML-gCMFD) iterative acceleration method and ML-gCMFD coupling acceleration method are proposed. In addition, the nonlinear preconditioning and hybrid perturbation size formula are implemented to further improve the convergence. Finally, to evaluate the numerical accuracy, convergence, efficiency and stability of MFNK method, a series of representative problems, including a three-dimensional (3D) single fuel pin problem, VERA Benchmark Problem 6, and VERA Benchmark Problem 7, are analyzed by comparing with the current N/TH coupling methods. Numerical results indicate that MFNK method can obtain strong stability, high convergence performance, and relatively high computational efficiency while ensuring high accuracy. It demonstrates that MFNK method has significant performance advantages and potential for high-fidelity N/TH coupling simulation.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207206","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 : 2024-09-06DOI: 10.1016/j.net.2024.09.006
Dana Hejazi, Anas M. Alwafi, Salman M. Alzahrani, Meshari M. Alqahtani, Salman M. Alshehri
Molten Salt Reactors (MSRs) represent a promising, safe, advanced nuclear technology that aligns with Saudi Arabia’s Vision 2030 objectives for energy diversification, sustainability, and economic development. This article explores the feasibility and potential impact of implementing MSRs in the Kingdom by assessing some of their technical, environmental, and socio-economic aspects. The research findings indicate that while MSRs may require higher initial capital investments, they offer substantial long-term socio-economic and environmental advantages over conventional power generation technologies. Economic advantages stem from improved fuel efficiency, lower waste management costs, fuel flexibility, load following, job creation, and operational cost savings. Environmental advantages include clean energy production, the ability to utilize thorium as a more environmentally friendly fuel source, the potential of nuclear transmutation for minimizing long-lived radioactive waste, and overall reduced waste output compared to traditional nuclear reactors. While implementing MSRs requires overcoming technical hurdles, mainly related to structural materials withstanding extreme conditions, Saudi Arabia’s thriving research ecosystem is well-suited for overcoming such challenges. The adoption of MSRs presents an opportunity for groundbreaking research, economic diversification, and progression towards a sustainable energy future. With rising energy demands and the need to transition towards cleaner sources, nuclear energy is poised to play a vital role globally and in Saudi Arabia’s future energy mix. This article hopes to raise awareness about the potential of advanced nuclear technologies like MSRs and inspire further research, investment, and conversation within the Kingdom to capitalize on this promising opportunity.
{"title":"The small modular molten salt reactor potential and opportunity in Saudi Arabia","authors":"Dana Hejazi, Anas M. Alwafi, Salman M. Alzahrani, Meshari M. Alqahtani, Salman M. Alshehri","doi":"10.1016/j.net.2024.09.006","DOIUrl":"https://doi.org/10.1016/j.net.2024.09.006","url":null,"abstract":"Molten Salt Reactors (MSRs) represent a promising, safe, advanced nuclear technology that aligns with Saudi Arabia’s Vision 2030 objectives for energy diversification, sustainability, and economic development. This article explores the feasibility and potential impact of implementing MSRs in the Kingdom by assessing some of their technical, environmental, and socio-economic aspects. The research findings indicate that while MSRs may require higher initial capital investments, they offer substantial long-term socio-economic and environmental advantages over conventional power generation technologies. Economic advantages stem from improved fuel efficiency, lower waste management costs, fuel flexibility, load following, job creation, and operational cost savings. Environmental advantages include clean energy production, the ability to utilize thorium as a more environmentally friendly fuel source, the potential of nuclear transmutation for minimizing long-lived radioactive waste, and overall reduced waste output compared to traditional nuclear reactors. While implementing MSRs requires overcoming technical hurdles, mainly related to structural materials withstanding extreme conditions, Saudi Arabia’s thriving research ecosystem is well-suited for overcoming such challenges. The adoption of MSRs presents an opportunity for groundbreaking research, economic diversification, and progression towards a sustainable energy future. With rising energy demands and the need to transition towards cleaner sources, nuclear energy is poised to play a vital role globally and in Saudi Arabia’s future energy mix. This article hopes to raise awareness about the potential of advanced nuclear technologies like MSRs and inspire further research, investment, and conversation within the Kingdom to capitalize on this promising opportunity.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142226263","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 : 2024-09-05DOI: 10.1016/j.net.2024.09.008
Xianzhe Duan, Cong Du, Nan Li, Jiaxin Duan, Zhenping Tang
Deep geological disposal is currently considered the most practical and feasible method for disposing high-level radioactive wastes (HLWs). One of its key scientific issues is the migration of nuclides in fissure media. However, studies on the migration of nuclides like U-238 are relatively limited. In this study, the granite rock masses in Beishan, Gansu were selected to construct a physical and mathematical model of U-238 migration using the Laplace transform and inverse transform methods. Based on the numerical methods, the kinetic migration of nuclide U-238 in the fissure media of granite was simulated, and the effects of parameters such as fissure width, hydraulic gradient, diffusible area ratio and rock porosity on the migration of U-238 were investigated. The following insights were obtained: (1) After 100,000 years, U-238 has a very limited diffusion depth in the matrix domain, with a diffusion range of only a few tens of millimeters, whereas it migrates relatively quickly in the fissure domain, with a maximum migration distance of about 1500 m. (2) Under the same migration time and distance, the relative concentration of nuclide U-238 in the fissure domain increases with larger gap width, hydraulic gradient, and diffusible area ratio, but decreases with higher rock porosity. (3) In the same time range, the rock masses with larger gap widths, hydraulic gradients, and diffusible area ratios have larger migration ranges, while those with higher porosities have smaller migration ranges. (4) While selecting a site for diposal of HLWs, it is recommended to choose rock masses in the granite area of Beishan with no fissures or few fissures; additionally, areas with smaller fissure widths, hydraulic gradients, and diffusible area ratios but higher rock porosity should be prioritized. This study can provide important theoretical support for understanding nuclide migration in the future geological disposal of HLWs.
{"title":"Kinetic simulation of uranium migration in granite fissure media of beishan, gansu, China: A case study based on the Laplace transform and inverse transform methods","authors":"Xianzhe Duan, Cong Du, Nan Li, Jiaxin Duan, Zhenping Tang","doi":"10.1016/j.net.2024.09.008","DOIUrl":"https://doi.org/10.1016/j.net.2024.09.008","url":null,"abstract":"Deep geological disposal is currently considered the most practical and feasible method for disposing high-level radioactive wastes (HLWs). One of its key scientific issues is the migration of nuclides in fissure media. However, studies on the migration of nuclides like U-238 are relatively limited. In this study, the granite rock masses in Beishan, Gansu were selected to construct a physical and mathematical model of U-238 migration using the Laplace transform and inverse transform methods. Based on the numerical methods, the kinetic migration of nuclide U-238 in the fissure media of granite was simulated, and the effects of parameters such as fissure width, hydraulic gradient, diffusible area ratio and rock porosity on the migration of U-238 were investigated. The following insights were obtained: (1) After 100,000 years, U-238 has a very limited diffusion depth in the matrix domain, with a diffusion range of only a few tens of millimeters, whereas it migrates relatively quickly in the fissure domain, with a maximum migration distance of about 1500 m. (2) Under the same migration time and distance, the relative concentration of nuclide U-238 in the fissure domain increases with larger gap width, hydraulic gradient, and diffusible area ratio, but decreases with higher rock porosity. (3) In the same time range, the rock masses with larger gap widths, hydraulic gradients, and diffusible area ratios have larger migration ranges, while those with higher porosities have smaller migration ranges. (4) While selecting a site for diposal of HLWs, it is recommended to choose rock masses in the granite area of Beishan with no fissures or few fissures; additionally, areas with smaller fissure widths, hydraulic gradients, and diffusible area ratios but higher rock porosity should be prioritized. This study can provide important theoretical support for understanding nuclide migration in the future geological disposal of HLWs.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142267046","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}
Fluence-to-dose conversion coefficients based on a developed rat model have been calculated for neutrons with energies <20 MeV using Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation was conducted respectively for 27 monodirectional monoenergetic neutron beams in the energy range 10 MeV to 20 MeV, under four different irradiation conditions: left lateral, right lateral, dorsal–ventral and ventral–dorsal. The neutron fluence-to-dose conversion coefficients for selected organs were presented in the study and can be used to determine the organ dose of the rats experimentally exposed to external neutron irradiation. The results in this work were compared with the published data based on a mouse model to investigate the effect of size and weight difference on neutron organ dose. The comparison results showed the fluence-to-dose conversion coefficients of the rat model have the similar energy dependency and sensitivity to irradiation conditions compared with that of the mouse model, and the weight and size difference in individuals could lead to different levels of neutron organ dose difference depending on neutron energy, irradiation conditions as well as the location of organs.
{"title":"Fluence-to-dose conversion coefficients in a voxel rat model for external neutron irradiation","authors":"Xiaomin Zhang, Xu Xu, Yong Yuan, Jing Ning, Dawei Li, Yunlong Ji","doi":"10.1016/j.net.2024.08.056","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.056","url":null,"abstract":"Fluence-to-dose conversion coefficients based on a developed rat model have been calculated for neutrons with energies <20 MeV using Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation was conducted respectively for 27 monodirectional monoenergetic neutron beams in the energy range 10 MeV to 20 MeV, under four different irradiation conditions: left lateral, right lateral, dorsal–ventral and ventral–dorsal. The neutron fluence-to-dose conversion coefficients for selected organs were presented in the study and can be used to determine the organ dose of the rats experimentally exposed to external neutron irradiation. The results in this work were compared with the published data based on a mouse model to investigate the effect of size and weight difference on neutron organ dose. The comparison results showed the fluence-to-dose conversion coefficients of the rat model have the similar energy dependency and sensitivity to irradiation conditions compared with that of the mouse model, and the weight and size difference in individuals could lead to different levels of neutron organ dose difference depending on neutron energy, irradiation conditions as well as the location of organs.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142226264","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 : 2024-09-03DOI: 10.1016/j.net.2024.08.042
Wanook Ji, Eunjoong Lee, Young-Yong Ji
High pressure ion chambers (HPIC) and NaI(Tl) scintillation detectors are widely used to monitor the ambient dose equivalent rate H*(10) within and around the Korean nuclear facilities. However, HPIC cannot provide spectrometric information and NaI(Tl) detector is limited in identifying nuclides, such as I, Cs, and Cs, released from nuclear facilities owing to its insufficient energy resolution. This study employed four halide scintillators – LaBr(Ce), CeBr, and SrI(Eu) – to measure the ambient dose equivalent rate and detect gamma nuclides from measured energy spectrum. First, the pulse–shaping time in the signal processing unit was optimized for each scintillator. Second, energy resolution and counting efficiency were estimated for Cs and Co. Finally, an irradiation test was performed to estimate the dose rate. Based on these results, LaBr(Ce) and NaI(Tl) were selected as in situ gamma spectrometry system for measuring environmental radiation, and field experiments were conducted near the Fukushima Daiichi nuclear power plant to measure the dose rate.
{"title":"Performances of halide scintillators for the dosimetry based on gamma-ray spectrometry for environmental monitoring systems","authors":"Wanook Ji, Eunjoong Lee, Young-Yong Ji","doi":"10.1016/j.net.2024.08.042","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.042","url":null,"abstract":"High pressure ion chambers (HPIC) and NaI(Tl) scintillation detectors are widely used to monitor the ambient dose equivalent rate H*(10) within and around the Korean nuclear facilities. However, HPIC cannot provide spectrometric information and NaI(Tl) detector is limited in identifying nuclides, such as I, Cs, and Cs, released from nuclear facilities owing to its insufficient energy resolution. This study employed four halide scintillators – LaBr(Ce), CeBr, and SrI(Eu) – to measure the ambient dose equivalent rate and detect gamma nuclides from measured energy spectrum. First, the pulse–shaping time in the signal processing unit was optimized for each scintillator. Second, energy resolution and counting efficiency were estimated for Cs and Co. Finally, an irradiation test was performed to estimate the dose rate. Based on these results, LaBr(Ce) and NaI(Tl) were selected as in situ gamma spectrometry system for measuring environmental radiation, and field experiments were conducted near the Fukushima Daiichi nuclear power plant to measure the dose rate.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207208","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 : 2024-09-02DOI: 10.1016/j.net.2024.08.063
Nassar Alnassar, Maha Algarawi, Sitah Alanazi, Muneerah Al-Aqeel, Ahmed Salah Khaliil, A. Abdelghafar Galahom
This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO, (Th, U)O, (Th, U)O and (Th, rgPu)O. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (k) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.
{"title":"Searching for an optimal fuel for a supercritical water reactor (SCWR) to dispose of minor actinides resulting from the nuclear waste","authors":"Nassar Alnassar, Maha Algarawi, Sitah Alanazi, Muneerah Al-Aqeel, Ahmed Salah Khaliil, A. Abdelghafar Galahom","doi":"10.1016/j.net.2024.08.063","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.063","url":null,"abstract":"This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO, (Th, U)O, (Th, U)O and (Th, rgPu)O. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (k) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207223","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 : 2024-09-02DOI: 10.1016/j.net.2024.08.060
S.Y. Zhang, Y.B. Nie, Y.Y. Ding, Q. Zhao, K.Z. Xu, X.Y. Pan, H.T. Chen, Q. Sun, Z. Wei
Integral experiments play a crucial role in advancing nuclear science and technology by providing critical data that validate theoretical models and enhance reactor designs. This study presents a novel approach to accurately determine pulse time distribution in integral experiments conducted with pulsed accelerators. By strategically placed monitors and shields at angles of 0° and 90° relative to the beam direction, neutron flight times from the target are measured, and a response matrix for neutron emission at different times is constructed through simulation. The Maximum Likelihood Expectation Maximization (MLEM) algorithm is employed for pulse time reconstruction, with the gamma ray flight time spectrum from monitors used as the initial spectrum to streamline the computational process. Experimental validation using a standard polyethylene sample and n-p scattering cross-sections confirms the accuracy of the method. Results are compared across multiple nuclear databases such as CENDL-3.2, ENDF/B-VIII.0, JENDL-5.0, and JEFF-3.3 libraries. The developed method significantly enhances the precision of pulse time distribution determination, thereby improving the quality and reliability of experimental data obtained from integral experiments conducted with pulsed accelerators.
{"title":"Accurate pulse time distribution determination using MLEM algorithm in integral experiments","authors":"S.Y. Zhang, Y.B. Nie, Y.Y. Ding, Q. Zhao, K.Z. Xu, X.Y. Pan, H.T. Chen, Q. Sun, Z. Wei","doi":"10.1016/j.net.2024.08.060","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.060","url":null,"abstract":"Integral experiments play a crucial role in advancing nuclear science and technology by providing critical data that validate theoretical models and enhance reactor designs. This study presents a novel approach to accurately determine pulse time distribution in integral experiments conducted with pulsed accelerators. By strategically placed monitors and shields at angles of 0° and 90° relative to the beam direction, neutron flight times from the target are measured, and a response matrix for neutron emission at different times is constructed through simulation. The Maximum Likelihood Expectation Maximization (MLEM) algorithm is employed for pulse time reconstruction, with the gamma ray flight time spectrum from monitors used as the initial spectrum to streamline the computational process. Experimental validation using a standard polyethylene sample and n-p scattering cross-sections confirms the accuracy of the method. Results are compared across multiple nuclear databases such as CENDL-3.2, ENDF/B-VIII.0, JENDL-5.0, and JEFF-3.3 libraries. The developed method significantly enhances the precision of pulse time distribution determination, thereby improving the quality and reliability of experimental data obtained from integral experiments conducted with pulsed accelerators.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207216","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 : 2024-09-02DOI: 10.1016/j.net.2024.08.065
S.F. Quintão Ribeiro, J.J. Rivero
The paper presents a model for operational risk of industrial facilities with applications to nuclear power plants and uranium mining. The traditional models update the risk profile for specific operational conditions setting to one the failure probability of components which are known to be unavailable by failure or maintenance, while constant mean unavailability values from the original Probabilistic Safety Analysis (PSA) are kept for the rest. The proposed methodology considers the time dependency of standby failure probabilities through instantaneous unavailability models, depending on the specific standby time for each component at the given moment, instead of using the same constant values. It also incorporates ageing, testing degradation effects and maintenance effectiveness, not explicitly considered in traditional models, as well as the instantaneous reevaluation of common cause failure probabilities. The results show significant risk underestimations when components specific standby times, ageing, testing degradation effects and maintenance effectiveness are not considered.
{"title":"A model for operational risk of industrial facilities including ageing/test degradation effects and maintenance effectiveness","authors":"S.F. Quintão Ribeiro, J.J. Rivero","doi":"10.1016/j.net.2024.08.065","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.065","url":null,"abstract":"The paper presents a model for operational risk of industrial facilities with applications to nuclear power plants and uranium mining. The traditional models update the risk profile for specific operational conditions setting to one the failure probability of components which are known to be unavailable by failure or maintenance, while constant mean unavailability values from the original Probabilistic Safety Analysis (PSA) are kept for the rest. The proposed methodology considers the time dependency of standby failure probabilities through instantaneous unavailability models, depending on the specific standby time for each component at the given moment, instead of using the same constant values. It also incorporates ageing, testing degradation effects and maintenance effectiveness, not explicitly considered in traditional models, as well as the instantaneous reevaluation of common cause failure probabilities. The results show significant risk underestimations when components specific standby times, ageing, testing degradation effects and maintenance effectiveness are not considered.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207209","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}
This paper reports the control strategy for the core power in an accelerator drive sub-critical (ADS) system. In an ADS system, the intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. The proposed control strategy is to control the reactor power by adjusting the proton beam power, where the beam power is adjusted by changing either the duty factor or the intensity of the proton beam. As an example, the reactor power control of the China initiative Accelerator Driven System (CiADS) facility has been studied by adjusting the beam power. Firstly, the beam power is set roughly by assigning a new duty factor, where the duty factor is set by changing the beam macro-pulse length and the pulse repetition rate of the proton beam. Both the pulse length and the repetition rate are assigned by a timing system. Secondly, the power is adjusted precisely by changing the beam intensity. To change continuously the beam intensity, an adjustable aperture is used to block the outer particles of the beam line in the accelerator. In order to evaluate the proposed control strategy, a CiADS core model is built based on the multi-node point reactor dynamics model. Three cases, the start of the facility, the decrease of core power and the increase of core power, have been simulated with the model. The simulation results indicate that the control strategy for the core power by changing either the duty factor or the intensity of the proton beam works very well during the operation of the facility.
{"title":"Control strategy for the core power in an accelerator drive sub-critical system","authors":"Xinxin Li, Yuan He, Wenjing Ma, Wenjuan Cui, Zhiyong He, Detai Zhou, Hai Zheng, Feng Yang, Yuhui Guo, Haihua Niu, Kai Yin, Shiwu Dang","doi":"10.1016/j.net.2024.08.066","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.066","url":null,"abstract":"This paper reports the control strategy for the core power in an accelerator drive sub-critical (ADS) system. In an ADS system, the intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. The proposed control strategy is to control the reactor power by adjusting the proton beam power, where the beam power is adjusted by changing either the duty factor or the intensity of the proton beam. As an example, the reactor power control of the China initiative Accelerator Driven System (CiADS) facility has been studied by adjusting the beam power. Firstly, the beam power is set roughly by assigning a new duty factor, where the duty factor is set by changing the beam macro-pulse length and the pulse repetition rate of the proton beam. Both the pulse length and the repetition rate are assigned by a timing system. Secondly, the power is adjusted precisely by changing the beam intensity. To change continuously the beam intensity, an adjustable aperture is used to block the outer particles of the beam line in the accelerator. In order to evaluate the proposed control strategy, a CiADS core model is built based on the multi-node point reactor dynamics model. Three cases, the start of the facility, the decrease of core power and the increase of core power, have been simulated with the model. The simulation results indicate that the control strategy for the core power by changing either the duty factor or the intensity of the proton beam works very well during the operation of the facility.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207215","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}
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