Pub Date : 2024-09-19DOI: 10.1016/j.net.2024.09.022
Mohamed Y. Hanfi , A.M. Abu El-Soad , Nadi Mlihan Alresheedi , Sultan J. Alsufyani , K.A. Mahmoud
The present study focuses on investigating the gamma-ray protection features of clay bricks for potential use in radiation shielding fields. The study examined the physical and structural features that affect the performance of these stones in shielding γ-rays. The density (ρ, g/cm3) of the clay bricks samples was measured utilizing the MH-300A density meter. Additionally, the mineral structure within the annealed pressed clay samples was identification the XRD spectrometry. Moreover, the morphology and elemental chemical composition for the annealed bricks were examined using a Thermo Scientific Prisma E, USA field emission Scanning Electron Microscope (SEM) in conjunction with Energy Dispersive X-ray Spectroscopy. Besides, the shielding features of the clay bricks were analyzed using the experimentally measurements (by NaI (Tl) scintillation detector), XCOM software, and Monte Carlo Simulation over the γ-ray energy interval of 0.033–1.332 MeV. The findings of the study indicate that an increase in the pressure rate within the clay bricks samples leads to the rise in their density (from 1.62 to 1.87 g/cm3). This increase in density is accompanied by a decline in both porosity (Φ, %) (from 34.75 to 26.21 %) and water absorption (K, %) (from 26.21 to 14.74 %) factors. Furthermore, the increase in pressure rate from 7.61 to 114.22 MPa also results in an increase in the linear attenuation coefficient (μ, cm−1) of the clay bricks under study. This is achieved by increasing the μ values from 0.39 to 0.43 cm−1, from 0.13 to 0.15 cm−1, and from 0.09 to 0.10 cm−1, at 0.081, 0.511 and 1.173 MeV, respectively. The synthetic bricks offer a lead-free and efficient option for protection, making them ideal for use in nuclear facility start-ups or in areas with radiation exposure.
{"title":"The impact of pressure rate on the physical, structural and gamma-ray shielding capabilities of novel light-weight clay bricks","authors":"Mohamed Y. Hanfi , A.M. Abu El-Soad , Nadi Mlihan Alresheedi , Sultan J. Alsufyani , K.A. Mahmoud","doi":"10.1016/j.net.2024.09.022","DOIUrl":"10.1016/j.net.2024.09.022","url":null,"abstract":"<div><div>The present study focuses on investigating the gamma-ray protection features of clay bricks for potential use in radiation shielding fields. The study examined the physical and structural features that affect the performance of these stones in shielding γ-rays. The density (ρ, g/cm<sup>3</sup>) of the clay bricks samples was measured utilizing the MH-300A density meter. Additionally, the mineral structure within the annealed pressed clay samples was identification the XRD spectrometry. Moreover, the morphology and elemental chemical composition for the annealed bricks were examined using a Thermo Scientific Prisma E, USA field emission Scanning Electron Microscope (SEM) in conjunction with Energy Dispersive X-ray Spectroscopy. Besides, the shielding features of the clay bricks were analyzed using the experimentally measurements (by NaI (Tl) scintillation detector), XCOM software, and Monte Carlo Simulation over the γ-ray energy interval of 0.033–1.332 MeV. The findings of the study indicate that an increase in the pressure rate within the clay bricks samples leads to the rise in their density (from 1.62 to 1.87 g/cm<sup>3</sup>). This increase in density is accompanied by a decline in both porosity (Φ, %) (from 34.75 to 26.21 %) and water absorption (K, %) (from 26.21 to 14.74 %) factors. Furthermore, the increase in pressure rate from 7.61 to 114.22 MPa also results in an increase in the linear attenuation coefficient (μ, cm<sup>−1</sup>) of the clay bricks under study. This is achieved by increasing the μ values from 0.39 to 0.43 cm<sup>−1</sup>, from 0.13 to 0.15 cm<sup>−1</sup>, and from 0.09 to 0.10 cm<sup>−1</sup>, at 0.081, 0.511 and 1.173 MeV, respectively. The synthetic bricks offer a lead-free and efficient option for protection, making them ideal for use in nuclear facility start-ups or in areas with radiation exposure.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 11","pages":"Pages 4938-4945"},"PeriodicalIF":2.6,"publicationDate":"2024-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142434256","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 : 2024-09-11DOI: 10.1016/j.net.2024.09.018
{"title":"Technical reviewers for Nuclear Engineering and Technology, 2024","authors":"","doi":"10.1016/j.net.2024.09.018","DOIUrl":"10.1016/j.net.2024.09.018","url":null,"abstract":"","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5437-5451"},"PeriodicalIF":2.6,"publicationDate":"2024-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142656863","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 : 2024-09-11DOI: 10.1016/j.net.2024.09.004
Sungho Moon , Haegin Han , Chansoo Choi , Bangho Shin , Gahee Son , Hyeonil Kim , Suhyeon Kim , Jaehyo Kim , In Gyu Yoon , Kyung Hwan Lee , Chan Hyeong Kim
The exposure devices utilizing high-activity gamma radiation sources for industrial radiography frequently encounter issues such as stuck or disconnected sources, posing a substantial risk of radiation exposure to the workers executing emergency source retrieval operations, emphasizing the importance of accurate dose assessment. In the present study, the radiation dose to the main worker during the process of source retrieval was calculated and compared for five emergency source retrieval procedures to retrieve stuck or disconnected sources, using 4D Monte Carlo dose calculation and a mesh-type reference computational phantom. For a stuck source, the dose values of the two source retrieval procedures (i.e., repair and cut methods) were found to be relatively small, which indicates that the worker might select either method based on personal preference or situational convenience. Conversely, for a disconnected source, the dose values of the three source retrieval procedures (i.e., fishing, connect/push, and hot stick methods) were much larger and show significant differences. Notably, the fishing method was associated with the lowest dose, whereas the hot stick method resulted in significantly higher doses, differences being as large as ∼5 times. These results underscore the fishing method as a preferable option, particularly over the hot stick method.
{"title":"Towards accurate dose assessment for emergency industrial radiography source retrieval operations: A preliminary study of 4D Monte Carlo dose calculations","authors":"Sungho Moon , Haegin Han , Chansoo Choi , Bangho Shin , Gahee Son , Hyeonil Kim , Suhyeon Kim , Jaehyo Kim , In Gyu Yoon , Kyung Hwan Lee , Chan Hyeong Kim","doi":"10.1016/j.net.2024.09.004","DOIUrl":"10.1016/j.net.2024.09.004","url":null,"abstract":"<div><div>The exposure devices utilizing high-activity gamma radiation sources for industrial radiography frequently encounter issues such as stuck or disconnected sources, posing a substantial risk of radiation exposure to the workers executing emergency source retrieval operations, emphasizing the importance of accurate dose assessment. In the present study, the radiation dose to the main worker during the process of source retrieval was calculated and compared for five emergency source retrieval procedures to retrieve stuck or disconnected sources, using 4D Monte Carlo dose calculation and a mesh-type reference computational phantom. For a stuck source, the dose values of the two source retrieval procedures (i.e., repair and cut methods) were found to be relatively small, which indicates that the worker might select either method based on personal preference or situational convenience. Conversely, for a disconnected source, the dose values of the three source retrieval procedures (i.e., fishing, connect/push, and hot stick methods) were much larger and show significant differences. Notably, the fishing method was associated with the lowest dose, whereas the hot stick method resulted in significantly higher doses, differences being as large as ∼5 times. These results underscore the fishing method as a preferable option, particularly over the hot stick method.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5428-5436"},"PeriodicalIF":2.6,"publicationDate":"2024-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142656380","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}
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":"7 1","pages":""},"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":"196 1","pages":""},"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":"11 1","pages":""},"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":"5 1","pages":""},"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":"316 1","pages":""},"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":"10 1","pages":""},"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.
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