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":null,"pages":null},"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":null,"pages":null},"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":null,"pages":null},"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":null,"pages":null},"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":null,"pages":null},"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":null,"pages":null},"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}
This paper discusses permanent deformation of the fuel assembly of pressurized water reactors that can satisfy a drop time requirement of the control rod assembly. During a seismic event, the fuel assemblies and the reactor core exhibit transient vibration. When the fuel assemblies collide, they develop impact forces in the spacer grids along the fuel assemblies. If the impact forces on some spacer grids of the fuel assembly exceed an allowable strength, the buckling occurs in the corresponding spacer grids, followed by permanent deformation. The geometry change of the buckled spacer grids leads to the movement of the guide tube positions in the fuel assembly. Therefore the geometry change is associated with the distortion of the control rod drop path. In this study, the permanent deformation of the spacer grids was measured during the impact tests, and the maximum deviation of the guide tube from the original position was reviewed. Control rod drop tests were also performed to measure the change in drop time according to the deformation of the fuel assembly. Using actuators in the loop test facility, the authors realized the first and second bending mode shapes of the fuel assembly under the flowing water condition. Based on a series of drop tests, the critical fuel assembly deformation was identified that could lead to a violation of an allowable drop time of the control rod.
{"title":"Experimental Investigation of Control Rod Drops into Perturbed Fuel Assemblies","authors":"N. Park, O. Kwon, Y. Yoo, Tae-Suk Kim","doi":"10.1115/1.4062275","DOIUrl":"https://doi.org/10.1115/1.4062275","url":null,"abstract":"\u0000 This paper discusses permanent deformation of the fuel assembly of pressurized water reactors that can satisfy a drop time requirement of the control rod assembly. During a seismic event, the fuel assemblies and the reactor core exhibit transient vibration. When the fuel assemblies collide, they develop impact forces in the spacer grids along the fuel assemblies. If the impact forces on some spacer grids of the fuel assembly exceed an allowable strength, the buckling occurs in the corresponding spacer grids, followed by permanent deformation. The geometry change of the buckled spacer grids leads to the movement of the guide tube positions in the fuel assembly. Therefore the geometry change is associated with the distortion of the control rod drop path. In this study, the permanent deformation of the spacer grids was measured during the impact tests, and the maximum deviation of the guide tube from the original position was reviewed. Control rod drop tests were also performed to measure the change in drop time according to the deformation of the fuel assembly. Using actuators in the loop test facility, the authors realized the first and second bending mode shapes of the fuel assembly under the flowing water condition. Based on a series of drop tests, the critical fuel assembly deformation was identified that could lead to a violation of an allowable drop time of the control rod.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-04-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83022679","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}
Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation.
{"title":"Preliminary Study on the Thermal Neutron Scattering Cross-Section for HinH2O in Small Modular Reactors","authors":"Jun Wu, Yixue Chen","doi":"10.3390/jne4020023","DOIUrl":"https://doi.org/10.3390/jne4020023","url":null,"abstract":"Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74838789","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}
We present measurement results for the flow rate of liquid sodium at temperatures up to 700°C that were obtained with a high temperature prototype of an immersed Eddy Current Flow Meter. The experimental campaign was conducted at the SOLTEC-2 sodium loop at KIT. The main objective of the experiments is the high temperature qualification of the Eddy Current Flow Meter as part of the safety instrumentation of generation IV liquid metal cooled fast reactors. There it is intended to be used for monitoring the flow rate of the coolant and to detect possible blockages of sub assemblies. Due to the large liquid metal volume, the sensor has to be located close to the sub assemblies, therefore measurements from outside of the vessel are not possible and an immersed sensor is required. We demonstrate the successful application of the immersed Eddy Current Flow Meter at such high temperatures and identify the relevant effects with impact on the sensor performance.
{"title":"Eddy Current Flow Meter Measurements in Liquid Sodium at High Temperatures","authors":"N. Krauter, A. Onea, G. Gerbeth, S. Eckert","doi":"10.1115/1.4062239","DOIUrl":"https://doi.org/10.1115/1.4062239","url":null,"abstract":"\u0000 We present measurement results for the flow rate of liquid sodium at temperatures up to 700°C that were obtained with a high temperature prototype of an immersed Eddy Current Flow Meter. The experimental campaign was conducted at the SOLTEC-2 sodium loop at KIT. The main objective of the experiments is the high temperature qualification of the Eddy Current Flow Meter as part of the safety instrumentation of generation IV liquid metal cooled fast reactors. There it is intended to be used for monitoring the flow rate of the coolant and to detect possible blockages of sub assemblies. Due to the large liquid metal volume, the sensor has to be located close to the sub assemblies, therefore measurements from outside of the vessel are not possible and an immersed sensor is required. We demonstrate the successful application of the immersed Eddy Current Flow Meter at such high temperatures and identify the relevant effects with impact on the sensor performance.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89229692","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}
J. Friconneau, T. Batal, O. David, Chiara Di Paolo, F. Ferlay, S. Gazzotti, L. Giancarli, Christophe Lacroix, J. Martins, Benjamin Michel, J. Travere
The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF.
{"title":"ITER Test Blanket Module—ALARA Investigations for Port Cell Pipe Forest Replacement","authors":"J. Friconneau, T. Batal, O. David, Chiara Di Paolo, F. Ferlay, S. Gazzotti, L. Giancarli, Christophe Lacroix, J. Martins, Benjamin Michel, J. Travere","doi":"10.3390/jne4010022","DOIUrl":"https://doi.org/10.3390/jne4010022","url":null,"abstract":"The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75411960","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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