Journal of Fusion Energy最新文献

{"title":"Upgrade of ITCD code and its Application to Global lithium Impurity Transport Modelling for EAST Tokamak","authors":"Y. L. Liu,&nbsp;Y. T. Chen,&nbsp;Z. H. Gao,&nbsp;C. W. Zhang,&nbsp;S. Y. Wang,&nbsp;S. Y. Dai","doi":"10.1007/s10894-025-00476-5","DOIUrl":"10.1007/s10894-025-00476-5","url":null,"abstract":"<div><p>Liquid metals, like lithium (Li), are considered a promising plasma-facing material due to their self-repairing, in comparison with the conventional solid materials that have limitations in handling high heat flux in future fusion devices. To predictively simulate global Li transport under the lithium divertor condition, the three-dimensional Monte Carlo code ITCD has been upgraded significantly, in terms of the simulation domain (from the sole divertor region in a limited toroidal range to the entire edge plasma region in a full toroidal torus). The expansion of the simulation zone brings about the new demand of the computational resource, which motivates us to implement the guiding-center (GC) particle push approach into ITCD. The trajectory of charged Li particle using the GC particle push approach shows a good agreement with the full-orbit (FO) particle push method. The FO and GC hybrid particle push scheme has been used to deal with the gyration scrape-off effect and meanwhile speed up the calculation of the global Li transport. The characteristics of Li impurity density and deposition distributions are studied in detail by ITCD.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 1","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143108134","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Exposure of Sn-Wetted W CPS Targets to Simultaneous NBI Beam and High-Power CW Laser Pulses at the High-Heat Flux OLMAT Facility","authors":"E. Oyarzabal,&nbsp;A. De Castro,&nbsp;D. Alegre,&nbsp;P. Fernandez-Mayo,&nbsp;D. Tafalla,&nbsp;K. J. McCarthy,&nbsp;The OLMAT Team","doi":"10.1007/s10894-025-00474-7","DOIUrl":"10.1007/s10894-025-00474-7","url":null,"abstract":"<div><p>First experiments are reported of the simultaneous exposure of a number of Sn-wetted W CPSs and a reference W CPS to 100 ms NBI pulses (divertor steady-state loading conditions) and 2 ms long high-energy laser pulses (divertor ELM like loading conditions) at the High-Heat Flux OLMAT facility. The use of a fast-frame imaging camera allows monitoring the onset of particle ejection from the targets during laser pulses and obtaining the corresponding laser heat fluxes as a measure of the resilience of these targets. Fast camera images are used also to determine ejected particle numbers and to estimate their maximum velocities as laser power is increased in order to compare the influence of W CPS structure on these parameters. In addition, the craters resulting from particle ejection are studied for each target with an optical microscope and a scanning electron microscope. Moreover, in-situ W and Sn particle ejection is followed using visible emission spectroscopy and post-exposure W melting after particle ejection is observed using the energy dispersive X-ray method EDX for all the studied targets. This shows that Sn is unable to protect the underlying W substrate from high-energy laser damage, albeit a subsequent refilling of the formed craters with Sn is visible during NBI-only pulses after laser damage. Thus, it is considered that optimization of surface refilling/replenishment with Sn is needed to improve the W substrate protection. From this work, it is also found that the W CPS reference material has a higher laser heat flux threshold for particle ejection than the Sn-wetted targets. Nevertheless, it is important to take into account that in these experiments with laser pulses, the possible beneficial effects of vapor shielding that can take place during particle irradiation at ELMs or disruptions are not present, thus these experiments represent a worst-case scenario.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 1","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-025-00474-7.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143109004","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Thermal Stability and Corrosion Resistance in Liquid Lithium of Brazed Tungsten Smart Alloy/RAFM Steel Joints","authors":"V. O. Kirillova,&nbsp;N. S. Popov,&nbsp;O. N. Sevryukov,&nbsp;X. Tan,&nbsp;A. A. Bazhenov,&nbsp;S. M. Irmagambetova,&nbsp;A. N. Suchkov","doi":"10.1007/s10894-024-00472-1","DOIUrl":"10.1007/s10894-024-00472-1","url":null,"abstract":"<div><p>Oxidation resistant smart tungsten alloys (SA) are considered a promising plasma facing material for DEMO reactors. SA-based plasma facing components (PFC) have to meet several long-term operation requirements. Among other criteria, these PFC should be able to withstand high thermal loads and be corrosion resistant in liquid lithium for a liquid first wall design implementation. In this work, smart tungsten alloys WCrY, WCrZr were brazed to reduced activation ferritic-martensitic (RAFM) steels Eurofer97, CLAM via 48Ti–48Zr–4Be wt.% brazing alloy. Thermal stability of the brazed joints was investigated. High temperature shear tests at 300, 600 °C were carried out. The shear strength of WCrZr/Ta/CLAM joints is 50 ± 4 and 49 ± 5 MPa at 300 and 600 °C, respectively. Unbrazing of the WCrY/Ta/Eurofer97 and WCrZr/Ta/CLAM joints occurs at 1447 and 1522 °C, respectively, due to the melting of steels. Corrosion resistance of the smart tungsten alloys, SA/Ta/RAFM joints in liquid lithium at 600 °C, 100 h exposure was demonstrated.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 1","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142798456","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Enhancements in Vertical Instability Control for the HL-3 Tokamak","authors":"Panle Liu,&nbsp;Bo Li,&nbsp;Xiang Chen,&nbsp;Shaoyong Liang,&nbsp;Qiang Li,&nbsp;Junzhao Zhang,&nbsp;Yihang Chen,&nbsp;Da Li","doi":"10.1007/s10894-024-00473-0","DOIUrl":"10.1007/s10894-024-00473-0","url":null,"abstract":"<div><p>Vertical position control of tokamak plasmas is essential for exploring operational limits and ensuring stable operation at high elongations to avoid disruptions. This study focuses on improving vertical instability control in the HL-3 tokamak by enhancing the signal-to-noise ratio of control signals and optimizing control strategies. We employed improved diagnostic techniques using Mirnov coils and flux loops, combined with digital filtering technology, to mitigate the effects of power supply switching and measurement noise. The vertical stabilization (VS) control system was upgraded with an optimized low-pass filter for vertical position estimation, a novel method for vertical velocity estimation using direct voltage signals from diagnostics, and an improved control algorithm. These enhancements resulted in significant improvements in control precision and noise reduction. Experimental results demonstrated successful control of highly elongated plasmas (<span>(kappa )</span> up to 1.8) with high plasma currents (up to 1.6 MA), achieving vertical position control accuracy better than 1 cm during the plasma current ramp-up phase. These advancements expand the operational parameter space of HL-3, paving the way for higher performance plasma operation.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 1","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142798455","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Preliminary Control-Oriented Modeling of the ITER Steering Mirror Assembly and Local Control System in the Electron Cyclotron Heating & Current Drive Actuator","authors":"G. Grapow,&nbsp;T. Ravensbergen,&nbsp;M. D’Onorio,&nbsp;F. Pesamosca,&nbsp;A. Vu,&nbsp;G. Carannante","doi":"10.1007/s10894-024-00465-0","DOIUrl":"10.1007/s10894-024-00465-0","url":null,"abstract":"<div><p>The ITER Electron Cyclotron Heating and Current Drive (ECH) plays a pivotal role in heating and controlling fusion plasmas, with the Steering Mirrors being a crucial component of this actuator. A representative model of the ECH is compulsory in the development and validation of the Plasma Control System (PCS). This manuscript aims to propose a Control-Oriented model of the Steering Mirrors based on the design tested at the Swiss Plasma Centre. In this design a steering mirror rotates on some frictionless flexure pivots due to the action of a set of externally pressurized bellows acting against pre-compressed springs. This system is referred to as the Steering Mirror Assembly (SMA). The adherence of the model is tested by comparing the simulations with the experimental results, while considering ITER’s most recent requirements. Performances, generally increased in terms of accuracy, are in line with the prototype’s results.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"43 2","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142540630","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research on Insulation Technology for Nb3Sn Layer Coil of Superconducting Conductor Testing Facility","authors":"Yuanyuan Ma,&nbsp;Bozhou Huang,&nbsp;Yongzheng Yao,&nbsp;Houxiang Han,&nbsp;Changneng Zhang,&nbsp;Lei Wu,&nbsp;Yi Shi","doi":"10.1007/s10894-024-00471-2","DOIUrl":"10.1007/s10894-024-00471-2","url":null,"abstract":"<div><p>The Superconducting Conductor Testing Facility, which is developed to evaluate the reliability of engineering technology and safe operation in a fusion reactor operation environment, is under construction by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Since the Nb₃Sn layer coil of this test facility adopts the manufacturing process of the wind &amp; react, the high-strength glass fiber used as the inter-turn insulation material will carbonize after high-temperature heat treatment at 665℃, thereby reducing the mechanical and electrical properties of the winding. The surface decarburization and modification process of high-strength glass fiber was developed to improve the properties of glass fiber after heat treatment. It is verified that the developed glass fiber tape treatment process can meet manufacturing process requirements of layered Nb₃Sn superconducting magnets through the coil winding radial pressure test and VPI sample performance test. This processing technology has been successfully applied in the manufacturing of experimental magnets, providing technical support for the insulation manufacturing of a large Nb₃Sn layer coil.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"43 2","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142524491","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"High Energy Density Radiative Transfer in the Diffusion Regime with Fourier Neural Operators","authors":"Joseph Farmer,&nbsp;Ethan Smith,&nbsp;William Bennett,&nbsp;Ryan McClarren","doi":"10.1007/s10894-024-00470-3","DOIUrl":"10.1007/s10894-024-00470-3","url":null,"abstract":"<div><p>Radiative heat transfer is a fundamental process in high energy density physics and inertial fusion. Accurately predicting the behavior of Marshak waves across a wide range of material properties and drive conditions is crucial for design and analysis of these systems. Conventional numerical solvers and analytical approximations often face challenges in terms of accuracy and computational efficiency. In this work, we propose a novel approach to model Marshak waves using Fourier Neural Operators (FNO). We develop two FNO-based models: (1) a base model that learns the mapping between the drive condition and material properties to a solution approximation based on the widely used analytic model by Hammer &amp; Rosen (2003), and (2) a model that corrects the inaccuracies of the analytic approximation by learning the mapping to a more accurate numerical solution. Our results demonstrate the strong generalization capabilities of the FNOs and show significant improvements in prediction accuracy compared to the base analytic model.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"43 2","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-024-00470-3.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142519003","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Retraction Note: Determination of the Plasma Internal Inductance and Evaluation of its Effects on Plasma Horizontal Displacement in IR-T1 Tokamak","authors":"A. Salar Elahi,&nbsp;M. Ghoranneviss","doi":"10.1007/s10894-024-00467-y","DOIUrl":"10.1007/s10894-024-00467-y","url":null,"abstract":"","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"43 2","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142438752","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Retraction Note: Differences Between the Toroidal and Poloidal Flux Loops in the Measurement of Plasma Position in Tokamaks","authors":"A. Salar Elahi,&nbsp;M. Ghoranneviss","doi":"10.1007/s10894-024-00466-z","DOIUrl":"10.1007/s10894-024-00466-z","url":null,"abstract":"","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"43 2","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-10-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142411518","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effects of Injected Current Streams on MHD Equilibrium Reconstruction of Local Helicity Injection Plasmas in a Spherical Tokamak","authors":"J. D. Weberski,&nbsp;M. W. Bongard,&nbsp;S. J. Diem,&nbsp;R. J. Fonck,&nbsp;J. A. Goetz,&nbsp;M. D. Nornberg,&nbsp;J. A. Reusch,&nbsp;C. E. Schaefer,&nbsp;A. C. Sontag","doi":"10.1007/s10894-024-00460-5","DOIUrl":"10.1007/s10894-024-00460-5","url":null,"abstract":"<div><p>Open field line currents are intrinsic to DC helicity injection plasma startup and pose a challenge for inferring the plasma equilibrium with standard reconstruction analysis. Local helicity injection (LHI) is a type of DC helicity injection which uses small, modular current sources to drive force-free current along helical field lines to produce tokamak plasmas. MHD modeling and magnetic measurements during LHI indicate the injected current streams remain coherent as helical structures on the outboard edge of a core toroidal plasma that is tokamak-like in a toroidally averaged sense. To extract core plasma equilibrium properties, external magnetic diagnostics corrected for contributions from the injected current streams are fitted by a standard Grad-Shafranov equilibrium code. An iterative approach for estimating and subtracting the stream contributions from the diagnostic signals is described and applied to a model equilibrium database to reduce systematic errors introduced by the streams. Convergence is usually attained with 2 to 4 iterations, with derived equilibrium parameters matching the prescribed axisymmetric core values to within estimated experimental uncertainties. Accurate recovery of core parameters occurs when the ratio of the net toroidal windup current from the streams to the core plasma current is less than 0.2, which is typically satisfied in most experiments.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"43 2","pages":""},"PeriodicalIF":1.9,"publicationDate":"2024-10-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-024-00460-5.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142411493","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}