Tu Nguyen, E. Patterson, Richard J. K. Taylor, Y. Tseng, C. Waldon
The differences between nuclear fission and fusion have been discussed widely in the literature. However, little has been done to investigate the key differences in safety designs and regulatory requirements between the nuclear reactor types. In this study, an innovative methodology was successfully developed to map nuclear safety features to the fundamental safety principles set out by the nuclear regulators. Three safety cases were assessed in the mapping study, a research fusion reactor (Joint European Torus), a research fission reactor (Tsing Hua Open-pool Reactor) and a commercial fission reactor (Hinkley Point C). The graphical representation allowed a comparative analysis of the safety features and fundamental principles which revealed differences between the hazard profiles of fission and fusion reactors and provided important insights for the creation of a similar map for a future commercial fusion device.
{"title":"Comparative Maps of Safety Features for Fission and Fusion Reactors","authors":"Tu Nguyen, E. Patterson, Richard J. K. Taylor, Y. Tseng, C. Waldon","doi":"10.1115/1.4062643","DOIUrl":"https://doi.org/10.1115/1.4062643","url":null,"abstract":"\u0000 The differences between nuclear fission and fusion have been discussed widely in the literature. However, little has been done to investigate the key differences in safety designs and regulatory requirements between the nuclear reactor types. In this study, an innovative methodology was successfully developed to map nuclear safety features to the fundamental safety principles set out by the nuclear regulators. Three safety cases were assessed in the mapping study, a research fusion reactor (Joint European Torus), a research fission reactor (Tsing Hua Open-pool Reactor) and a commercial fission reactor (Hinkley Point C). The graphical representation allowed a comparative analysis of the safety features and fundamental principles which revealed differences between the hazard profiles of fission and fusion reactors and provided important insights for the creation of a similar map for a future commercial fusion device.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72454669","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Small Modular Reactors (SMRs) have gained international attention due to their modular design, small footprint, and lower capital costs for research, development, and construction compared to conventional reactors. Many types of SMRs are being developed in different countries, and regulatory agencies are working on a robust and “harmonized” SMR regulatory framework to ensure safety and protect the environment. However, there are still many details that need to be understood, such as nuclear fuel behavior at high pressures and temperatures (1000 °C), types and levels of radiation exposure during normal operations and accidents, types and volume of nuclear waste, and their proper storage and disposal. Moreover, SMRs' modular design and small size make them suitable for remote locations, such as the Canadian Arctic. However, before introducing this technology, a detailed study of the arctic soil (permafrost) is needed in the context of changing climate. Probabilistic risk assessment (PRA) is a crucial methodology for assessing the safety and reliability of nuclear power plants. Due to multi-module nature of SMRs, cross-unit interactions ( multi-module effects) need to be evaluated as part of the total plant safety assessment. Additionally, given the nature of fuels (low-enriched uranium) and the possible remote location with minimal technical staff, nuclear materials may have a higher probability of diversion. Therefore, nuclear safeguards and material accountancy are essential to prevent nuclear proliferation. This paper discusses the benefits and challenges of deploying different SMR technologies for electricity generation and other applications.
{"title":"Small Modular Reactors: Opportunities and Challenges as Emerging Nuclear Technologies for Power Production","authors":"L. Ghimire, E. Waller","doi":"10.1115/1.4062644","DOIUrl":"https://doi.org/10.1115/1.4062644","url":null,"abstract":"\u0000 Small Modular Reactors (SMRs) have gained international attention due to their modular design, small footprint, and lower capital costs for research, development, and construction compared to conventional reactors. Many types of SMRs are being developed in different countries, and regulatory agencies are working on a robust and “harmonized” SMR regulatory framework to ensure safety and protect the environment. However, there are still many details that need to be understood, such as nuclear fuel behavior at high pressures and temperatures (1000 °C), types and levels of radiation exposure during normal operations and accidents, types and volume of nuclear waste, and their proper storage and disposal. Moreover, SMRs' modular design and small size make them suitable for remote locations, such as the Canadian Arctic. However, before introducing this technology, a detailed study of the arctic soil (permafrost) is needed in the context of changing climate. Probabilistic risk assessment (PRA) is a crucial methodology for assessing the safety and reliability of nuclear power plants. Due to multi-module nature of SMRs, cross-unit interactions ( multi-module effects) need to be evaluated as part of the total plant safety assessment. Additionally, given the nature of fuels (low-enriched uranium) and the possible remote location with minimal technical staff, nuclear materials may have a higher probability of diversion. Therefore, nuclear safeguards and material accountancy are essential to prevent nuclear proliferation. This paper discusses the benefits and challenges of deploying different SMR technologies for electricity generation and other applications.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"32 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87101671","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pierrick Agullo, A. Gossard, G. Ranchoux, E. Porcheron, Fabrice Petitot
Several nuclear facilities are currently dismantling in France, namely on CEA's and EDF's sites. The whole decommissioning and dismantling, cutting operations and nuclear waste management can significantly affect worker risks. In this context, this work aims to better assess the internal exposure risks for operations where removal factor and airborne release factor play an essential part in the uncertainties. For that purpose, developing a new method for assessing the risk of internal exposure is necessary to optimize the choice of Personal Protective Equipment during nuclear dismantling. Based on a bibliographic study and feedback, we identified most of the forces and parameters that influence labile contamination removal and particles resuspension. Their high number forced us to highlight the most relevant ones regarding data from the bibliography and worksite reality. Thus, the role of the drying temperature, relative humidity, roughness of the contaminated surface and the swiping pressure were particularly studied on the assessment of the labile contamination and the removal factor. For that, we performed successive wiping operations with cotton wipes on surface on which a simulated labile contamination was deposited to estimate the influence of different parameters on the total labile contamination and the removal factor. Thus, we have highlighted the high influence of roughness on these latter compare to the other identified parameters.
{"title":"Main Parameters Influencing The Level Of Labile Contamination And The Removal Factor","authors":"Pierrick Agullo, A. Gossard, G. Ranchoux, E. Porcheron, Fabrice Petitot","doi":"10.1115/1.4062550","DOIUrl":"https://doi.org/10.1115/1.4062550","url":null,"abstract":"\u0000 Several nuclear facilities are currently dismantling in France, namely on CEA's and EDF's sites. The whole decommissioning and dismantling, cutting operations and nuclear waste management can significantly affect worker risks. In this context, this work aims to better assess the internal exposure risks for operations where removal factor and airborne release factor play an essential part in the uncertainties. For that purpose, developing a new method for assessing the risk of internal exposure is necessary to optimize the choice of Personal Protective Equipment during nuclear dismantling.\u0000 Based on a bibliographic study and feedback, we identified most of the forces and parameters that influence labile contamination removal and particles resuspension. Their high number forced us to highlight the most relevant ones regarding data from the bibliography and worksite reality. Thus, the role of the drying temperature, relative humidity, roughness of the contaminated surface and the swiping pressure were particularly studied on the assessment of the labile contamination and the removal factor. For that, we performed successive wiping operations with cotton wipes on surface on which a simulated labile contamination was deposited to estimate the influence of different parameters on the total labile contamination and the removal factor. Thus, we have highlighted the high influence of roughness on these latter compare to the other identified parameters.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"81 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76852060","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Nicholas Dunkle, Alex Wheeler, Jarod Richardson, Sandra Bogetic, Ondrej Chvala, Steven E. Skutnik
Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion.
{"title":"Plutonium Signatures in Molten-Salt Reactor Off-Gas Tank and Safeguards Considerations","authors":"Nicholas Dunkle, Alex Wheeler, Jarod Richardson, Sandra Bogetic, Ondrej Chvala, Steven E. Skutnik","doi":"10.3390/jne4020028","DOIUrl":"https://doi.org/10.3390/jne4020028","url":null,"abstract":"Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"109 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135718117","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A system thermal-hydraulics model for a fluoride-salt-cooled high-temperature reactor (FHR) based on the small modular advanced high-temperature reactor (SmAHTR) design concept is developed, using RELAP5-3D. The SmAHTR components modelled in the simulations include: the reactor core, lower plenum, upper plenum, top plenum, three Primary Heat Exchangers (PHX's) equipped with three primary pumps, and three Director Reactor Auxiliary Cooling System (DRACS) equipped with three fluid diodes. Flows through the reactor core are represented by 19 individual fuel channels, one reflector-hole channel, and one downcomer channel. In each of the 19 SmAHTR fuel block channels, the fuel elements are modeled in 5 groups using 5 heat structures, each with their corresponding power level. The total reactor power is 125 MWth. Using representative core power distributions for the SmAHTR at beginning-of-cycle (BOC) and at end-of-cycle (EOC), two steady-state system thermal-hydraulic model simulations with RELAP5-3D were performed using a default pressure drop loss factor of 1.5 for all 19 fuel channels. Exit coolant temperatures ranged from 688°C to 739°C (BOC) and from 696°C to 721°C (EOC), while peak fuel centerline temperatures in the highest power block were 1,249°C (BOC) and 1,029°C (EOC). By adjusting the loss factors to modify coolant flow rates in each channel, a more uniform exit coolant temperature was possible.
{"title":"System Thermal-hydraulics Model for Fluoride Salt-Cooled Reactor Based On Small Advanced High Temperature Reactor (smAHTR) Design Concept","authors":"Shu Jun Wang, Xianmin Huang, B. Bromley","doi":"10.1115/1.4062500","DOIUrl":"https://doi.org/10.1115/1.4062500","url":null,"abstract":"\u0000 A system thermal-hydraulics model for a fluoride-salt-cooled high-temperature reactor (FHR) based on the small modular advanced high-temperature reactor (SmAHTR) design concept is developed, using RELAP5-3D. The SmAHTR components modelled in the simulations include: the reactor core, lower plenum, upper plenum, top plenum, three Primary Heat Exchangers (PHX's) equipped with three primary pumps, and three Director Reactor Auxiliary Cooling System (DRACS) equipped with three fluid diodes. Flows through the reactor core are represented by 19 individual fuel channels, one reflector-hole channel, and one downcomer channel. In each of the 19 SmAHTR fuel block channels, the fuel elements are modeled in 5 groups using 5 heat structures, each with their corresponding power level. The total reactor power is 125 MWth. Using representative core power distributions for the SmAHTR at beginning-of-cycle (BOC) and at end-of-cycle (EOC), two steady-state system thermal-hydraulic model simulations with RELAP5-3D were performed using a default pressure drop loss factor of 1.5 for all 19 fuel channels. Exit coolant temperatures ranged from 688°C to 739°C (BOC) and from 696°C to 721°C (EOC), while peak fuel centerline temperatures in the highest power block were 1,249°C (BOC) and 1,029°C (EOC). By adjusting the loss factors to modify coolant flow rates in each channel, a more uniform exit coolant temperature was possible.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"140 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85405420","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer material within which most of the subsequent deformation is localized. Channeling is often associated with hypothetical dislocation pileup and intergranular cracking in reactor components although the channels themselves do not amplify stress as one would expect from a pileup. The channels are often similar in appearance to twins leading to the possibility that twins are sometimes mistakenly identified as channels. Neither twins nor dislocation channels, which are bulk shears, produce the same stress conditions as a pileup on a single plane. At high doses, when cavities are produced (either He-stabilised bubbles at low temperatures or voids at high temperatures), there can be reduced ductility because the material is already in an equivalent advanced stage of microscopic necking. He-stabilised cavities form preferentially on grain boundaries and at precipitate or incoherent twin/ε-martensite interfaces. The higher planar density of the cavities, coupled with the incompatibility at the interface, results in a preferential failure known as He embrittlement. Strain localisation and inter- or intragranular failure are dependent on many factors that are ultimately microstructural in nature. The mechanisms are described and discussed in relation to reactor core materials.
{"title":"Strain Localisation and Fracture of Nuclear Reactor Core Materials","authors":"M. Griffiths","doi":"10.3390/jne4020026","DOIUrl":"https://doi.org/10.3390/jne4020026","url":null,"abstract":"The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer material within which most of the subsequent deformation is localized. Channeling is often associated with hypothetical dislocation pileup and intergranular cracking in reactor components although the channels themselves do not amplify stress as one would expect from a pileup. The channels are often similar in appearance to twins leading to the possibility that twins are sometimes mistakenly identified as channels. Neither twins nor dislocation channels, which are bulk shears, produce the same stress conditions as a pileup on a single plane. At high doses, when cavities are produced (either He-stabilised bubbles at low temperatures or voids at high temperatures), there can be reduced ductility because the material is already in an equivalent advanced stage of microscopic necking. He-stabilised cavities form preferentially on grain boundaries and at precipitate or incoherent twin/ε-martensite interfaces. The higher planar density of the cavities, coupled with the incompatibility at the interface, results in a preferential failure known as He embrittlement. Strain localisation and inter- or intragranular failure are dependent on many factors that are ultimately microstructural in nature. The mechanisms are described and discussed in relation to reactor core materials.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"5 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76735181","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Alexander Lau, J. Coenen, D. Schwalenberg, Y. Mao, T. Höschen, Johann Riesch, L. Raumann, Michael Treitz, Hanns Gietl, A. Terra, Beatrix Göhts, C. Linsmeier, K. Theis-Bröhl, J. Gonzalez‐Julian
The use of tungsten fiber-reinforced tungsten composites (Wf/W) has been demonstrated to significantly enhance the mechanical properties of tungsten (W) by incorporating W-fibers into the W-matrix. However, prior research has been restricted by the usage of single fiber-based textile fabrics, consisting of 150 µm warp and 50 µm weft filaments, with limited homogeneity, reproducibility, and mechanical properties in bulk structures due to the rigidity of the 150 µm W-fibers. To overcome this limitation, two novel textile preforms were developed utilizing radial braided W-yarns with 7 core and 16 sleeve filaments (R.B. 16 + 7), with a diameter of 25 µm each, as the warp material. In this study, bulk composites of two different fabric types were produced via a layer-by-layer CVD process, utilizing single 50 µm filaments (type 1) and R.B. 16 + 7 yarns (type 2) as weft materials. The produced composites were sectioned into KLST-type specimens based on DIN EN ISO 179-1:2000 using electrical discharge machining (EDM) and subjected to three-point bending tests. Both composites demonstrated enhanced mechanical properties with pseudo-ductile behavior at room temperature and withstood over 10,000 load cycles between 50–90% of their respective maximum load without sample fracture in three-point cyclic loading tests. Furthermore, a novel approach to predict the fatigue behavior of the material under cyclic loading was developed based on the high reproducibility of the composites produced, especially for the composite based on type 1. This approach provides a new benchmark for upscaling endeavors and may enable a better prediction of the service life of the produced components made of Wf/W in the future. In comparison, the composite based on fabric type 1 demonstrated superior results in manufacturing performance and mechanical properties. With a high relative average density (>97%), a high fiber volume fraction (14–17%), and a very homogeneous fiber distribution in the CVD-W matrix, type 1 shows a promising option to be further tested in high heat flux tests and to be potentially used as an alternative to currently used materials for the most stressed components of nuclear fusion reactors or other potential application fields such as concentrated solar power (CSP), aircraft turbines, the steel industry, quantum computing, or welding tools. Type 2 composites have a higher layer spacing compared to type 1, resulting in gaps within the matrix and less homogeneous material properties. While type 2 composites have demonstrated a notable enhancement over 150 µm fiber-based composites, they are not viable for industrial scale-up unlike type 1 composites.
钨纤维增强钨复合材料(Wf/W)通过在钨基体中加入钨纤维，可以显著提高钨的力学性能。然而，先前的研究受到单一纤维基纺织织物的限制，由150µm经纱和50µm纬纱组成，由于150µm w -纤维的刚性，在大块结构中均匀性、再现性和机械性能有限。为了克服这一限制，研究人员开发了两种新型纺织预成型材料，采用径向编织w纱，其中含有7芯和16袖长丝(r.b.16 + 7)，每根直径为25微米。在本研究中，使用单50微米长丝(类型1)和rb 16 + 7纱线(类型2)作为纬纱材料，通过逐层CVD工艺生产了两种不同织物类型的大块复合材料。根据DIN EN ISO 179-1:2000使用电火花加工(EDM)将生产的复合材料切割成klst型试样，并进行三点弯曲试验。两种复合材料在室温下均表现出增强的力学性能和伪延性，并在三点循环加载试验中承受了超过10,000次载荷循环，载荷范围为各自最大载荷的50-90%，无试样断裂。此外，基于所生产的复合材料，特别是基于类型1的复合材料的高重复性，开发了一种新的方法来预测材料在循环载荷下的疲劳行为。这种方法为升级工作提供了一个新的基准，并且可以更好地预测未来由Wf/W制成的生产部件的使用寿命。相比之下，基于织物类型1的复合材料在制造性能和机械性能方面表现出更优越的结果。1型具有较高的相对平均密度(>97%)、较高的纤维体积分数(14-17%)和CVD-W基体中非常均匀的纤维分布，显示出在高热流密度测试中进一步测试的有希望的选择，并有可能作为目前使用的材料的替代品用于核聚变反应堆的最大应力部件或其他潜在的应用领域，如聚光太阳能(CSP)、飞机涡轮机、钢铁工业、量子计算。或焊接工具。与1型相比，2型复合材料具有更高的层间距，导致基体内部存在间隙，材料性能不均匀。虽然2型复合材料在超过150 μ m的纤维基复合材料中表现出显著的增强，但与1型复合材料不同，它们不适合工业规模。
{"title":"Bulk Tungsten Fiber-Reinforced Tungsten (Wf/W) Composites Using Yarn-Based Textile Preforms","authors":"Alexander Lau, J. Coenen, D. Schwalenberg, Y. Mao, T. Höschen, Johann Riesch, L. Raumann, Michael Treitz, Hanns Gietl, A. Terra, Beatrix Göhts, C. Linsmeier, K. Theis-Bröhl, J. Gonzalez‐Julian","doi":"10.3390/jne4020027","DOIUrl":"https://doi.org/10.3390/jne4020027","url":null,"abstract":"The use of tungsten fiber-reinforced tungsten composites (Wf/W) has been demonstrated to significantly enhance the mechanical properties of tungsten (W) by incorporating W-fibers into the W-matrix. However, prior research has been restricted by the usage of single fiber-based textile fabrics, consisting of 150 µm warp and 50 µm weft filaments, with limited homogeneity, reproducibility, and mechanical properties in bulk structures due to the rigidity of the 150 µm W-fibers. To overcome this limitation, two novel textile preforms were developed utilizing radial braided W-yarns with 7 core and 16 sleeve filaments (R.B. 16 + 7), with a diameter of 25 µm each, as the warp material. In this study, bulk composites of two different fabric types were produced via a layer-by-layer CVD process, utilizing single 50 µm filaments (type 1) and R.B. 16 + 7 yarns (type 2) as weft materials. The produced composites were sectioned into KLST-type specimens based on DIN EN ISO 179-1:2000 using electrical discharge machining (EDM) and subjected to three-point bending tests. Both composites demonstrated enhanced mechanical properties with pseudo-ductile behavior at room temperature and withstood over 10,000 load cycles between 50–90% of their respective maximum load without sample fracture in three-point cyclic loading tests. Furthermore, a novel approach to predict the fatigue behavior of the material under cyclic loading was developed based on the high reproducibility of the composites produced, especially for the composite based on type 1. This approach provides a new benchmark for upscaling endeavors and may enable a better prediction of the service life of the produced components made of Wf/W in the future. In comparison, the composite based on fabric type 1 demonstrated superior results in manufacturing performance and mechanical properties. With a high relative average density (>97%), a high fiber volume fraction (14–17%), and a very homogeneous fiber distribution in the CVD-W matrix, type 1 shows a promising option to be further tested in high heat flux tests and to be potentially used as an alternative to currently used materials for the most stressed components of nuclear fusion reactors or other potential application fields such as concentrated solar power (CSP), aircraft turbines, the steel industry, quantum computing, or welding tools. Type 2 composites have a higher layer spacing compared to type 1, resulting in gaps within the matrix and less homogeneous material properties. While type 2 composites have demonstrated a notable enhancement over 150 µm fiber-based composites, they are not viable for industrial scale-up unlike type 1 composites.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"42 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73794827","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This second edition of the Handbook of Generation IV Nuclear Reactors (https://www.elsevier.com/books/handbook-of-generation-iv-nuclear-reactors/9780128205884) combines the history of development, research, industrial-operating experience, advanced designs, systems and safety analysis, and applications of nuclear energy. Impressive number of 64 recognized nuclear-energy-system experts from 13 nuclear-power countries: Belgium, Canada, China, France, Germany, India, Japan, Russia, Republic of Korea, Ukraine, Switzerland, UK, and USA, contributed to this book. The second edition builds on the successes of the first edition published in 2016 (Handbook of Generation IV Nuclear Reactors, 2016. Editor: I.L. Pioro, Elsevier - Woodhead Publishing).
{"title":"Handbook of Generation IV Nuclear Reactors Edition 2","authors":"J. Riznic","doi":"10.1115/1.4062402","DOIUrl":"https://doi.org/10.1115/1.4062402","url":null,"abstract":"\u0000 This second edition of the Handbook of Generation IV Nuclear Reactors (https://www.elsevier.com/books/handbook-of-generation-iv-nuclear-reactors/9780128205884) combines the history of development, research, industrial-operating experience, advanced designs, systems and safety analysis, and applications of nuclear energy. Impressive number of 64 recognized nuclear-energy-system experts from 13 nuclear-power countries: Belgium, Canada, China, France, Germany, India, Japan, Russia, Republic of Korea, Ukraine, Switzerland, UK, and USA, contributed to this book. The second edition builds on the successes of the first edition published in 2016 (Handbook of Generation IV Nuclear Reactors, 2016. Editor: I.L. Pioro, Elsevier - Woodhead Publishing).","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"11 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-04-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77012371","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Mohsen Hosseinzadehtaher, Silvanus D’silva, Matthew Baker, Ritesh Kumar, Nathan T. Hein, M. Shadmand, S. Jagadish, B. Ghanbarian
This article studies the optimal design approach for a portable nuclear magnetic resonance (NMR) system for use in non-destructive flow measurement applications. The mechanical and electromagnetic design procedures were carried out using the Ansys Maxwell finite-element analysis (FEA) software tool. The proposed procedure considered homogeneity and strength constraints while ensuring the desired functionality of the intended device for a given application. A modified particle swarm optimization (MPSO) algorithm was proposed as a reference design framework for optimization stages. The optimally designed NMR tool was prototyped, and its functionality was validated via several case studies. To assess the functionality of the prototyped device, Larmor frequency for hydrogen atom was captured and compared with theoretical results. Furthermore, the functionality and accuracy of the prototyped NMR tool is compared to the off-the-shelf NMR tool. Results demonstrated the feasibility and accuracy of the prototyped NMR tool constrained by factors, such as being lightweight and compact.
{"title":"On Design Challenges of Portable Nuclear Magnetic Resonance System","authors":"Mohsen Hosseinzadehtaher, Silvanus D’silva, Matthew Baker, Ritesh Kumar, Nathan T. Hein, M. Shadmand, S. Jagadish, B. Ghanbarian","doi":"10.3390/jne4020025","DOIUrl":"https://doi.org/10.3390/jne4020025","url":null,"abstract":"This article studies the optimal design approach for a portable nuclear magnetic resonance (NMR) system for use in non-destructive flow measurement applications. The mechanical and electromagnetic design procedures were carried out using the Ansys Maxwell finite-element analysis (FEA) software tool. The proposed procedure considered homogeneity and strength constraints while ensuring the desired functionality of the intended device for a given application. A modified particle swarm optimization (MPSO) algorithm was proposed as a reference design framework for optimization stages. The optimally designed NMR tool was prototyped, and its functionality was validated via several case studies. To assess the functionality of the prototyped device, Larmor frequency for hydrogen atom was captured and compared with theoretical results. Furthermore, the functionality and accuracy of the prototyped NMR tool is compared to the off-the-shelf NMR tool. Results demonstrated the feasibility and accuracy of the prototyped NMR tool constrained by factors, such as being lightweight and compact.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"34 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-04-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87012065","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Machine learning methods allow the prediction of material properties, potentially using only the elemental composition of a molecule or compound, without the knowledge of molecular or crystalline structures. Herein, a composition-based machine learning prediction of the material properties of V–Cr–Ti alloys is demonstrated. Our machine-learning-based prediction of the stability of the V–Cr–Ti alloys is qualitatively consistent with the composition-dependent experimental data of the ductile–brittle transition temperature and swelling. Furthermore, our computational results suggest the existence of a composition region, Cr+Ti ~ 60 wt.%, at a significantly low ductile–brittle transition temperature. This outcome contrasts with a reportedly low Cr+Ti content of less than 10 wt.% in conventional V–Cr–Ti alloys. Machine-learning-based numerical stability prediction is useful for the design and analysis of metal alloys, particularly for multicomponent alloys such as high-entropy alloys, to develop materials for nuclear fusion reactors.
{"title":"Machine-Learning-Based Composition Analysis of the Stability of V–Cr–Ti Alloys","authors":"K. Tanabe","doi":"10.3390/jne4020024","DOIUrl":"https://doi.org/10.3390/jne4020024","url":null,"abstract":"Machine learning methods allow the prediction of material properties, potentially using only the elemental composition of a molecule or compound, without the knowledge of molecular or crystalline structures. Herein, a composition-based machine learning prediction of the material properties of V–Cr–Ti alloys is demonstrated. Our machine-learning-based prediction of the stability of the V–Cr–Ti alloys is qualitatively consistent with the composition-dependent experimental data of the ductile–brittle transition temperature and swelling. Furthermore, our computational results suggest the existence of a composition region, Cr+Ti ~ 60 wt.%, at a significantly low ductile–brittle transition temperature. This outcome contrasts with a reportedly low Cr+Ti content of less than 10 wt.% in conventional V–Cr–Ti alloys. Machine-learning-based numerical stability prediction is useful for the design and analysis of metal alloys, particularly for multicomponent alloys such as high-entropy alloys, to develop materials for nuclear fusion reactors.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"29 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83075560","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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