Pub Date : 2024-07-05DOI: 10.1016/j.net.2024.07.014
M. Lovecký, J. Závorka
The paper explores the management of spent nuclear fuel, focusing on dual-purpose spent fuel casks used to store and transport spent nuclear fuel from VVER-440 reactors. The main goal is to optimize spent fuel cask loading by developing an extensive methodology supported by a powerful tool. Using a multiple-zoning strategy, cooler outside fuel assemblies protect radiation sources from the hotter inner assemblies. An effective tool based on adjoint particle flux calculations is the recently developed OPOS-440 calculation code. This code allows for optimizing the loading pattern and determining dose rates surrounding the spent nuclear fuel cask for a selected fuel loading. The code also thoroughly demonstrates how different fuel assemblies affect the dose rate. These findings have real-world implications for reactor operations, including optimizing cask loading and supporting the licensing procedure for novel fuel types in already-existing spent fuel casks.
{"title":"Optimizing spent nuclear fuel cask loading for VVER-440 fuel","authors":"M. Lovecký, J. Závorka","doi":"10.1016/j.net.2024.07.014","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.014","url":null,"abstract":"The paper explores the management of spent nuclear fuel, focusing on dual-purpose spent fuel casks used to store and transport spent nuclear fuel from VVER-440 reactors. The main goal is to optimize spent fuel cask loading by developing an extensive methodology supported by a powerful tool. Using a multiple-zoning strategy, cooler outside fuel assemblies protect radiation sources from the hotter inner assemblies. An effective tool based on adjoint particle flux calculations is the recently developed OPOS-440 calculation code. This code allows for optimizing the loading pattern and determining dose rates surrounding the spent nuclear fuel cask for a selected fuel loading. The code also thoroughly demonstrates how different fuel assemblies affect the dose rate. These findings have real-world implications for reactor operations, including optimizing cask loading and supporting the licensing procedure for novel fuel types in already-existing spent fuel casks.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612794","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 : 2024-07-04DOI: 10.1016/j.net.2024.07.008
Ayhan Yılmazer, Gökhan Pediz
Effects of relative motion between moving fuel and neutrons are usually not considered in the analysis of Circulating Fuel Reactors (CFRs). In this study, we formulate neutron transport equation for CFRs in a hydrodynamic representation in terms of velocities relative to moving fuel. Using the P1 approximation in the comoving transport equation, the diffusion equation for CFRs is obtained. Mass transport of precursors is considered in the formulations. Comoving frame CFRs equations are analytically solved for critical slab problem and a closed form criticality condition is obtained. Comoving representation has introduced corrections to Eulerian cross sections arising from the acceleration term. Hydrodynamics coupling has posed density modifications to cross sections. A parametric study of corrective terms is carried out to calculate effects of these corrections on a generic MSR and on Molten Salt Breeder Reactor (MSBR).
{"title":"Hydrodynamics coupled circulating-fuel reactor equations in comoving frame and their analytical solutions for molten salt reactors","authors":"Ayhan Yılmazer, Gökhan Pediz","doi":"10.1016/j.net.2024.07.008","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.008","url":null,"abstract":"Effects of relative motion between moving fuel and neutrons are usually not considered in the analysis of Circulating Fuel Reactors (CFRs). In this study, we formulate neutron transport equation for CFRs in a hydrodynamic representation in terms of velocities relative to moving fuel. Using the P1 approximation in the comoving transport equation, the diffusion equation for CFRs is obtained. Mass transport of precursors is considered in the formulations. Comoving frame CFRs equations are analytically solved for critical slab problem and a closed form criticality condition is obtained. Comoving representation has introduced corrections to Eulerian cross sections arising from the acceleration term. Hydrodynamics coupling has posed density modifications to cross sections. A parametric study of corrective terms is carried out to calculate effects of these corrections on a generic MSR and on Molten Salt Breeder Reactor (MSBR).","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612797","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 : 2024-07-04DOI: 10.1016/j.net.2024.07.007
João D. Talon, Aquilino S. Martinez, Alessandro C. Gonçalves
Monitoring the radial power distribution during the operation of a pressurized light water reactor (PWR) is crucial for ensuring safe operating conditions and achieving high levels of fuel burnup. This paper introduces a methodology utilizing Artificial Neural Networks (ANN) for reconstructing the radial power distribution in the core of a Pressurized Water Reactor (PWR) with a hot full power of 1876 MWth, such as the Angra 1 reactor. This approach uses measurements from Self-Powered Neutron Detectors (SPND), simulated through the SERPENT code. The use of ANN demonstrated good accuracy in predicting the radial power distribution with an average relative error of 1.27%, considering 36 active detectors, with maximum relative error of 6.99%. Moreover, the proposed process demonstrated robust performance, even when measurements from one, two, or three SPND detectors were unavailable, resulting in errors of 1.24%, 1.13 %, and 1.09%, respectively. Therefore, the methodology ensures a reliable reconstruction of the radial power distribution, even when SPND detector measurements are unavailable, enabling the optimization of detector use and contributing to the increase of operational safety margins.
{"title":"Continuous mapping of nuclear reactor core power using artificial neural network even in the presence of inactive detectors","authors":"João D. Talon, Aquilino S. Martinez, Alessandro C. Gonçalves","doi":"10.1016/j.net.2024.07.007","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.007","url":null,"abstract":"Monitoring the radial power distribution during the operation of a pressurized light water reactor (PWR) is crucial for ensuring safe operating conditions and achieving high levels of fuel burnup. This paper introduces a methodology utilizing Artificial Neural Networks (ANN) for reconstructing the radial power distribution in the core of a Pressurized Water Reactor (PWR) with a hot full power of 1876 MWth, such as the Angra 1 reactor. This approach uses measurements from Self-Powered Neutron Detectors (SPND), simulated through the SERPENT code. The use of ANN demonstrated good accuracy in predicting the radial power distribution with an average relative error of 1.27%, considering 36 active detectors, with maximum relative error of 6.99%. Moreover, the proposed process demonstrated robust performance, even when measurements from one, two, or three SPND detectors were unavailable, resulting in errors of 1.24%, 1.13 %, and 1.09%, respectively. Therefore, the methodology ensures a reliable reconstruction of the radial power distribution, even when SPND detector measurements are unavailable, enabling the optimization of detector use and contributing to the increase of operational safety margins.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612798","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 : 2024-07-04DOI: 10.1016/j.net.2024.07.009
Xiaoliang Zou, Yanting Sun, Bo Yang, Yibao Liu
Zirconium hydride (ZrH) is an ideal moderator for space nuclear reactors due to its exceptional properties, including high hydrogen content, small neutron absorption cross section, and high operating temperature. The primary goal of this study is to examine how the atomic ratio of H to Zr (H/Zr ratio) influences the neutronics parameters of a ZrH moderated space nuclear reactor, with a focus on establishing a reliable reference for ensuring the optimal safety and minimization of such reactors. The neutronics calculation based on the ZrH moderated space nuclear reactor named Topaz-II is performed using the Reactor Monte Carlo code (RMC code) with the ENDF/VII cross-section database. The effects of the H/Zr ratio were studied with a particular focus on the initial k, the burnup, the temperature reactivity coefficient and the criticality safety. The results show that with the increase of the H/Zr ratio, the initial k increases while the drums’ worth decreases. The Moderator Temperature Coefficient (MTC) is a positive value that rises as the H/Zr ratio increases. In the dropping accidents, the reactor full of voids with seawater is more serious, and the introduced reactivity decreases with the increase of the H/Zr ratio.
{"title":"Analysis of the atomic ratio of H and Zr effect on the neutronics parameters of ZrH moderated space nuclear reactor","authors":"Xiaoliang Zou, Yanting Sun, Bo Yang, Yibao Liu","doi":"10.1016/j.net.2024.07.009","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.009","url":null,"abstract":"Zirconium hydride (ZrH) is an ideal moderator for space nuclear reactors due to its exceptional properties, including high hydrogen content, small neutron absorption cross section, and high operating temperature. The primary goal of this study is to examine how the atomic ratio of H to Zr (H/Zr ratio) influences the neutronics parameters of a ZrH moderated space nuclear reactor, with a focus on establishing a reliable reference for ensuring the optimal safety and minimization of such reactors. The neutronics calculation based on the ZrH moderated space nuclear reactor named Topaz-II is performed using the Reactor Monte Carlo code (RMC code) with the ENDF/VII cross-section database. The effects of the H/Zr ratio were studied with a particular focus on the initial k, the burnup, the temperature reactivity coefficient and the criticality safety. The results show that with the increase of the H/Zr ratio, the initial k increases while the drums’ worth decreases. The Moderator Temperature Coefficient (MTC) is a positive value that rises as the H/Zr ratio increases. In the dropping accidents, the reactor full of voids with seawater is more serious, and the introduced reactivity decreases with the increase of the H/Zr ratio.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612795","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 : 2024-07-04DOI: 10.1016/j.net.2024.07.013
M.I. Sayyed, M. Rashad, Chaitali V. More, Anjan Kumar
This work assessed highly effective bismuth borosilicate glasses' nuclear radiation shielding capabilities in terms of different concentrations of BiO. Melt-quenching was the method used to create the glasses. The Phy-X/PSD program was used to compute the μ values. The bismuth borosilicate glasses' nuclear radiation shielding properties were determined in the 0.015–15 MeV energy range. Increasing Bi₂O₃ concentration resulted in an increase in mass attenuation coefficient, linear attenuation coefficient, and effective atomic number. Conversely, the half value layer and mean free path values decreased. The B48Bi17 sample, with its largest Z, is a great gamma attenuator. The results of a new study may help to clarify the nature of the BiO additive used in borosilicate glasses, which are a potential form of shield for use in industrial and medical radiation facilities.
{"title":"Preparation and radiation shielding features of high density, transparent borosilicate glasses with different Bi2O3 contents","authors":"M.I. Sayyed, M. Rashad, Chaitali V. More, Anjan Kumar","doi":"10.1016/j.net.2024.07.013","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.013","url":null,"abstract":"This work assessed highly effective bismuth borosilicate glasses' nuclear radiation shielding capabilities in terms of different concentrations of BiO. Melt-quenching was the method used to create the glasses. The Phy-X/PSD program was used to compute the μ values. The bismuth borosilicate glasses' nuclear radiation shielding properties were determined in the 0.015–15 MeV energy range. Increasing Bi₂O₃ concentration resulted in an increase in mass attenuation coefficient, linear attenuation coefficient, and effective atomic number. Conversely, the half value layer and mean free path values decreased. The B48Bi17 sample, with its largest Z, is a great gamma attenuator. The results of a new study may help to clarify the nature of the BiO additive used in borosilicate glasses, which are a potential form of shield for use in industrial and medical radiation facilities.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612799","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 : 2024-07-03DOI: 10.1016/j.net.2024.06.031
This study analyzes the single-event transient (SET) characteristics of alpha particles on multi-channel Forksheet-FET and Nanosheet-FET at the device and circuit levels through 3D TCAD simulations. The study investigates the differences in SET responses based on the energy and incident position of incoming alpha particles, considering the structural variances between Forksheet-FET and Nanosheet-FET, as well as the presence or absence of bottom dielectric isolation (BDI) in the fabrication process. Specifically, the introduction of BDI is observed to significantly suppress the voltage drop caused by ‘unintended' current, as it can block the substantial electron-hole pairs (EHP) generated by injected alpha particles in the bulk substrate from reaching the FET terminals. Furthermore, it was confirmed that the size of abnormal current decreases as the energy of the injected alpha particle increases. Additionally, evaluating the response to SET based on the fundamental logic circuit, the CMOS inverter, revealed relatively small abnormal voltage drops for both Forksheet and Nanosheet when BDI was applied, confirming high immunity to radiation effects. Moreover, it can be observed that the application of BDI enhances reliability from a memory perspective by effectively suppressing voltage flips in the SRAM's cross-coupled latch circuit.
本研究通过三维 TCAD 仿真，在器件和电路层面分析了α粒子对多通道叉片场效应晶体管和纳米片场效应晶体管的单次瞬态（SET）特性。考虑到 Forksheet-FET 和 Nanosheet-FET 在结构上的差异，以及在制造过程中是否存在底部介质隔离 (BDI)，该研究根据入射α粒子的能量和入射位置，研究了 SET 响应的差异。具体来说，引入 BDI 可以显著抑制由 "意外 "电流引起的电压降，因为它可以阻止由注入的α粒子在块状衬底中产生的大量电子-空穴对（EHP）到达 FET 端子。此外，研究还证实，异常电流的大小会随着注入α粒子能量的增加而减小。此外，根据基本逻辑电路（CMOS 逆变器）评估对 SET 的响应时发现，在应用 BDI 时，Forksheet 和 Nanosheet 的异常压降都相对较小，这证实了对辐射影响的高度免疫性。此外，从存储器的角度来看，应用 BDI 可以有效抑制 SRAM 交叉耦合锁存电路中的电压翻转，从而提高可靠性。
{"title":"Radiation effects on multi-channel Forksheet-FET and Nanosheet-FET considering the bottom dielectric isolation scheme","authors":"","doi":"10.1016/j.net.2024.06.031","DOIUrl":"10.1016/j.net.2024.06.031","url":null,"abstract":"<div><div>This study analyzes the single-event transient (SET) characteristics of alpha particles on multi-channel Forksheet-FET and Nanosheet-FET at the device and circuit levels through 3D TCAD simulations. The study investigates the differences in SET responses based on the energy and incident position of incoming alpha particles, considering the structural variances between Forksheet-FET and Nanosheet-FET, as well as the presence or absence of bottom dielectric isolation (BDI) in the fabrication process. Specifically, the introduction of BDI is observed to significantly suppress the voltage drop caused by ‘unintended' current, as it can block the substantial electron-hole pairs (EHP) generated by injected alpha particles in the bulk substrate from reaching the FET terminals. Furthermore, it was confirmed that the size of abnormal current decreases as the energy of the injected alpha particle increases. Additionally, evaluating the response to SET based on the fundamental logic circuit, the CMOS inverter, revealed relatively small abnormal voltage drops for both Forksheet and Nanosheet when BDI was applied, confirming high immunity to radiation effects. Moreover, it can be observed that the application of BDI enhances reliability from a memory perspective by effectively suppressing voltage flips in the SRAM's cross-coupled latch circuit.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.6,"publicationDate":"2024-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612806","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 : 2024-07-03DOI: 10.1016/j.net.2024.07.006
Daeyoun Lee, Woo J. Kim, Young Rok Choi
This study investigates the phenomenon of confirmation bias in information selection in the context of nuclear spent fuel (NSF) storage facilities. An online survey was administered to a sample of 321 residents in South Korea. They were asked to assess their beliefs about the safety/risk of NSF storage facilities before and after their exposure to additional information. Our findings show a positive association between the initial belief and confirmation bias, suggesting that the research participants tend to select articles consistent with their beliefs about the safety/risk of the facility. Trust in government is negatively related to confirmation bias, implying that residents with a greater level of trust in government are more likely to choose information opposing their initial beliefs. Finally, this study finds the self-reinforcing and potentially polarizing nature of individuals' beliefs about the safety/risk of NSF storage facilities as residents’ initial beliefs are reinforced rather than challenged after they reviewed self-selected articles. Implications for public policy and communication strategies related to improving the acceptance of NSF storage facilities are discussed.
{"title":"Confirmation bias, information selection, and belief reinforcement about the safety/risk of nuclear spent fuel storage facilities","authors":"Daeyoun Lee, Woo J. Kim, Young Rok Choi","doi":"10.1016/j.net.2024.07.006","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.006","url":null,"abstract":"This study investigates the phenomenon of confirmation bias in information selection in the context of nuclear spent fuel (NSF) storage facilities. An online survey was administered to a sample of 321 residents in South Korea. They were asked to assess their beliefs about the safety/risk of NSF storage facilities before and after their exposure to additional information. Our findings show a positive association between the initial belief and confirmation bias, suggesting that the research participants tend to select articles consistent with their beliefs about the safety/risk of the facility. Trust in government is negatively related to confirmation bias, implying that residents with a greater level of trust in government are more likely to choose information opposing their initial beliefs. Finally, this study finds the self-reinforcing and potentially polarizing nature of individuals' beliefs about the safety/risk of NSF storage facilities as residents’ initial beliefs are reinforced rather than challenged after they reviewed self-selected articles. Implications for public policy and communication strategies related to improving the acceptance of NSF storage facilities are discussed.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612810","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 : 2024-07-03DOI: 10.1016/j.net.2024.06.048
Realistic simulation of nuclear fuel performance requires not only validated models capable of describing the thermomechanical phenomena that take place within the fuel under irradiation conditions, but a detailed description of the thermal hydraulics of the channel surrounding the fuel rods, which provides the boundary conditions of the system. In this work, the main results and outlooks of coupling the thermal hydraulics code SubChanFlow with the fuel performance code DIONISIO are presented. To achieve this, an internal coupling was implemented, wherein DIONISIO is used as a master code controlling SubChanFlow as a thermal hydraulics subroutine replacing the simplified version already embedded in DIONISIO. Several tests were conducted to ensure the performance and quality of the coupling under normal operation conditions as a first approach. In addition, it was observed that the coupling demonstrated a significant improvement in the description of the cladding temperature and related variables, such as oxide thickness and hydrogen uptake, when compared with experimental data.
{"title":"Assessing the impact of DIONISIO-SubChanFlow code coupling in nuclear fuel performance simulations","authors":"","doi":"10.1016/j.net.2024.06.048","DOIUrl":"10.1016/j.net.2024.06.048","url":null,"abstract":"<div><div>Realistic simulation of nuclear fuel performance requires not only validated models capable of describing the thermomechanical phenomena that take place within the fuel under irradiation conditions, but a detailed description of the thermal hydraulics of the channel surrounding the fuel rods, which provides the boundary conditions of the system. In this work, the main results and outlooks of coupling the thermal hydraulics code SubChanFlow with the fuel performance code DIONISIO are presented. To achieve this, an internal coupling was implemented, wherein DIONISIO is used as a master code controlling SubChanFlow as a thermal hydraulics subroutine replacing the simplified version already embedded in DIONISIO. Several tests were conducted to ensure the performance and quality of the coupling under normal operation conditions as a first approach. In addition, it was observed that the coupling demonstrated a significant improvement in the description of the cladding temperature and related variables, such as oxide thickness and hydrogen uptake, when compared with experimental data.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.6,"publicationDate":"2024-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612800","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 : 2024-07-02DOI: 10.1016/j.net.2024.07.001
Medical radioisotopes (RIs) are widely used in the diagnosis and treatment of tumors. Theranostic RIs, such as 131I, 166Ho, 177Lu, and 186Re, are particularly notable for their ability to enable diagnosis and treatment simultaneously. Conducting irradiation tests using the High-flux Advanced Neutron Application Reactor (HANARO), a 30-MW multipurpose research reactor, enables the production of these medical RIs. Prior to irradiation tests, it is crucial to assess whether they affect the thermal stability of irradiated materials and the thermal-hydraulic safety of HANARO. This study systematically investigates the feasibility of 166Ho production using the isotope production irradiation hole of HANARO, focusing on thermal stability. The nuclear heating rates of the RI production target and RI capsule are calculated using MCNP6, and the calculated nuclear heating rates are used for three-dimensional heat transfer analysis using COMSOL Multiphysics. Under the assumed conditions in this study, 166Ho production did not compromise the thermal stability of the RI production target and RI capsule; consequently, the thermal-hydraulic safety criteria of HANARO could be satisfied. This study can serve as a valuable reference for evaluating the thermal stability of irradiated materials and the thermal-hydraulic safety of HANARO, which must be performed before irradiation tests using HANARO.
医用放射性同位素（RIs）被广泛用于肿瘤的诊断和治疗。Theranostic RIs，如 I、Ho、Lu 和 Re，因其能够同时进行诊断和治疗而特别引人注目。利用高通量先进中子应用反应堆（HANARO）（30 兆瓦的多用途研究反应堆）进行辐照试验，可以生产这些医用 RIs。在进行辐照试验之前，评估这些试验是否会影响辐照材料的热稳定性以及 HANARO 的热液压安全性至关重要。本研究系统地研究了利用 HANARO 的同位素生产辐照孔生产 Ho 的可行性，重点关注热稳定性。使用 MCNP6 计算了 RI 生产靶和 RI 胶囊的核加热率，并使用 COMSOL Multiphysics 对计算出的核加热率进行了三维传热分析。在本研究假定的条件下，Ho 生产不会影响 RI 生产靶和 RI 胶囊的热稳定性，因此可以满足 HANARO 的热-水力安全标准。本研究可为评估辐照材料的热稳定性和 HANARO 的热液压安全性提供有价值的参考。
{"title":"Systematic analysis for the thermal stability assessment of 166Ho production using HANARO: An in silico study","authors":"","doi":"10.1016/j.net.2024.07.001","DOIUrl":"10.1016/j.net.2024.07.001","url":null,"abstract":"<div><div>Medical radioisotopes (RIs) are widely used in the diagnosis and treatment of tumors. Theranostic RIs, such as <sup>131</sup>I, <sup>166</sup>Ho, <sup>177</sup>Lu, and <sup>186</sup>Re, are particularly notable for their ability to enable diagnosis and treatment simultaneously. Conducting irradiation tests using the High-flux Advanced Neutron Application Reactor (HANARO), a 30-MW multipurpose research reactor, enables the production of these medical RIs. Prior to irradiation tests, it is crucial to assess whether they affect the thermal stability of irradiated materials and the thermal-hydraulic safety of HANARO. This study systematically investigates the feasibility of <sup>166</sup>Ho production using the isotope production irradiation hole of HANARO, focusing on thermal stability. The nuclear heating rates of the RI production target and RI capsule are calculated using MCNP6, and the calculated nuclear heating rates are used for three-dimensional heat transfer analysis using COMSOL Multiphysics. Under the assumed conditions in this study, <sup>166</sup>Ho production did not compromise the thermal stability of the RI production target and RI capsule; consequently, the thermal-hydraulic safety criteria of HANARO could be satisfied. This study can serve as a valuable reference for evaluating the thermal stability of irradiated materials and the thermal-hydraulic safety of HANARO, which must be performed before irradiation tests using HANARO.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.6,"publicationDate":"2024-07-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612803","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 : 2024-07-02DOI: 10.1016/j.net.2024.06.019
The current study has focused on evolving materials with high attenuation performance and vigorous mechanical properties. Nano hexagonal boron nitride (h-BN) has been synthesized from boric acid and urea and then heated at a low temperature. XRD, SEM, EDX-mapping images, and FTIR investigated the nano h-BN synthesized and boron oxide (B2O3) milled. The new nanocomposites based on ethylene propylene diene rubber (EPDM), different concentrations of h-BN NPs, and B2O3 have been prepared. The physical, mechanical, and thermal properties and neutron attenuation behavior of nanocomposites were characterized. The shielding properties were determined by measuring the fast neutrons and total gamma-ray attenuation coefficients of the 239Pu-α-9Be neutron source. It was seen that adding a coupling agent (maleic anhydride) was appropriate for improving interfacial adhesions between EPDM chains and the nanofiller h-BN and B2O3 compared to EPDM unloaded. From the results, we observe that EPDM loaded with nano h-BN and nano h-BN/B2O3 was noticeably boosted over that of the unloaded EPDM for tensile strength (TS), EPDM/nano h-BN/B2O3 had a high TS at a concentration of 3 % h-BN/10 % B2O3 (42 %) compared with unloaded EPDM. Moreover, EPDM/3 % h-BN and EPDM/3 % h-BN/10 % B2O3 had the highest thermal stability until 490 °C compared to unloaded EPDM is stable at 350 °C. Finally, the maximum macroscopic effective shielding behaviors and removal cross-section are estimated by incorporating h-BN 3 % into EPDM.
{"title":"Neutron attenuation performance of EPDM rubber with BN Nanoparticles/B2O3 composite and studying physical, thermal and mechanical properties","authors":"","doi":"10.1016/j.net.2024.06.019","DOIUrl":"10.1016/j.net.2024.06.019","url":null,"abstract":"<div><div>The current study has focused on evolving materials with high attenuation performance and vigorous mechanical properties. Nano hexagonal boron nitride (h-BN) has been synthesized from boric acid and urea and then heated at a low temperature. XRD, SEM, EDX-mapping images, and FTIR investigated the nano h-BN synthesized and boron oxide (B2O3) milled. The new nanocomposites based on ethylene propylene diene rubber (EPDM), different concentrations of h-BN NPs, and B<sub>2</sub>O<sub>3</sub> have been prepared. The physical, mechanical, and thermal properties and neutron attenuation behavior of nanocomposites were characterized. The shielding properties were determined by measuring the fast neutrons and total gamma-ray attenuation coefficients of the <sup>239</sup>Pu-α-<sup>9</sup>Be neutron source. It was seen that adding a coupling agent (maleic anhydride) was appropriate for improving interfacial adhesions between EPDM chains and the nanofiller h-BN and B<sub>2</sub>O<sub>3</sub> compared to EPDM unloaded. From the results, we observe that EPDM loaded with nano h-BN and nano h-BN/B<sub>2</sub>O<sub>3</sub> was noticeably boosted over that of the unloaded EPDM for tensile strength (TS), EPDM/nano h-BN/B<sub>2</sub>O<sub>3</sub> had a high TS at a concentration of 3 % h-BN/10 % B<sub>2</sub>O<sub>3</sub> (42 %) compared with unloaded EPDM. Moreover, EPDM/3 % h-BN and EPDM/3 % h-BN/10 % B<sub>2</sub>O<sub>3</sub> had the highest thermal stability until 490 °C compared to unloaded EPDM is stable at 350 °C. Finally, the maximum macroscopic effective shielding behaviors and removal cross-section are estimated by incorporating h-BN 3 % into EPDM.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.6,"publicationDate":"2024-07-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141698021","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}
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