Pub Date : 2024-11-13DOI: 10.1016/j.fusengdes.2024.114726
Zhenhan Li , Hua Li , Cunwen Tang , Guanghong Wang , Xiaohua Bao , Ge Gao
Quench protection system (QPS) is critical for ensuring the safety operation of superconducting magnets in nuclear fusion devices, while the pyro-breaker is wildly used in QPS for backup protection. This paper discusses the feasibility of using Underwater Electrical Wire Explosion (UEWE) to drive pyro-breaker. Firstly, the overall structure and the driving mechanism of arc contact of pyro-breaker are analyzed. Secondly, the detonation pressure required for contact breaking is measured by experiment. Then a microsecond-level UEWE experimental platform was established to test the discharge characteristics and shock wave intensity during the electrical explosion process. Finally, the detonation pressure generated by UEWE is measured and compared with contact driving mechanism of pyro-breaker. The experiment demonstrates that the UEWE has sufficient potential for replacing the explosive and driving the pyro-breaker. This study provides a new pathway for the design of the pyro-breaker.
{"title":"Preliminary study on electrical wire explosion utilized in pyro-breaker","authors":"Zhenhan Li , Hua Li , Cunwen Tang , Guanghong Wang , Xiaohua Bao , Ge Gao","doi":"10.1016/j.fusengdes.2024.114726","DOIUrl":"10.1016/j.fusengdes.2024.114726","url":null,"abstract":"<div><div>Quench protection system (QPS) is critical for ensuring the safety operation of superconducting magnets in nuclear fusion devices, while the pyro-breaker is wildly used in QPS for backup protection. This paper discusses the feasibility of using Underwater Electrical Wire Explosion (UEWE) to drive pyro-breaker. Firstly, the overall structure and the driving mechanism of arc contact of pyro-breaker are analyzed. Secondly, the detonation pressure required for contact breaking is measured by experiment. Then a microsecond-level UEWE experimental platform was established to test the discharge characteristics and shock wave intensity during the electrical explosion process. Finally, the detonation pressure generated by UEWE is measured and compared with contact driving mechanism of pyro-breaker. The experiment demonstrates that the UEWE has sufficient potential for replacing the explosive and driving the pyro-breaker. This study provides a new pathway for the design of the pyro-breaker.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114726"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142655584","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In ITER, the first mirrors of optical diagnostics will be prone to deposition of the first wall materials which will result in the degradation of the mirror's optical parameters and the parameters of the diagnostic itself. To restore the mirror's reflectivity a cleaning system based on the gas discharge will be employed. The present paper discussed low-frequency power modulation of capacitively-coupled RF discharge utilized to decrease the mean RF power supplied through the in-vessel feeding line. The impact of power modulation on the incident ion flux, distribution of ion energy, and sputtering rates of the first wall materials were studied for the cases of DC-coupled and -decoupled mirrors with/without external magnetic field. The parameters of the plasma were measured depending on modulation frequency (1 Hz – 10 kHz) and duty cycle (10 – 90 %) for He plasma at a pressure of a few Pa. Based on the measurements of ion flux and ion energy the sputtered rates of Be and BeO were estimated. The estimations of the sputtering rate were followed by experiments on BeO sputtering in He plasma. Finally, the recommendations are given for selecting an appropriate RF power required to clean the mirror from Be/BeO deposits in ITER.
在热核实验堆中，光学诊断仪的第一面镜子很容易沉积第一壁材料，导致镜子的光学参数和诊断仪本身的参数降低。为了恢复反射镜的反射率，将采用基于气体放电的清洁系统。本文讨论了利用电容耦合射频放电的低频功率调制来降低通过容器内馈电线提供的平均射频功率。针对有/无外部磁场的直流耦合镜和去耦合镜，研究了功率调制对入射离子通量、离子能量分布和第一壁材料溅射率的影响。根据调制频率（1 Hz - 10 kHz）和占空比（10 - 90 %），在几帕压力下测量了 He 等离子体的参数。根据离子通量和离子能量的测量结果，估算出 Be 和 BeO 的溅射率。在对溅射率进行估算之后，还对 He 等离子体中的 BeO 溅射进行了实验。最后，就选择适当的射频功率以清除热核实验堆中反射镜上的铍/氧化铍沉积物提出了建议。
{"title":"Pulsed radiofrequency plasma for cleaning ITER first mirrors with and without notch-filter and magnetic field","authors":"A.M. Dmitriev , A.G. Razdobarin , L.A. Snigirev , D.I. Elets , I.M. Bukreev , E.E. Mukhin , S.Yu. Tolstyakov , I.B. Kupriyanov , L. Moser","doi":"10.1016/j.fusengdes.2024.114724","DOIUrl":"10.1016/j.fusengdes.2024.114724","url":null,"abstract":"<div><div>In ITER, the first mirrors of optical diagnostics will be prone to deposition of the first wall materials which will result in the degradation of the mirror's optical parameters and the parameters of the diagnostic itself. To restore the mirror's reflectivity a cleaning system based on the gas discharge will be employed. The present paper discussed low-frequency power modulation of capacitively-coupled RF discharge utilized to decrease the mean RF power supplied through the in-vessel feeding line. The impact of power modulation on the incident ion flux, distribution of ion energy, and sputtering rates of the first wall materials were studied for the cases of DC-coupled and -decoupled mirrors with/without external magnetic field. The parameters of the plasma were measured depending on modulation frequency (1 Hz – 10 kHz) and duty cycle (10 – 90 %) for He plasma at a pressure of a few Pa. Based on the measurements of ion flux and ion energy the sputtered rates of Be and BeO were estimated. The estimations of the sputtering rate were followed by experiments on BeO sputtering in He plasma. Finally, the recommendations are given for selecting an appropriate RF power required to clean the mirror from Be/BeO deposits in ITER.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114724"},"PeriodicalIF":1.9,"publicationDate":"2024-11-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142655781","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-02DOI: 10.1016/j.fusengdes.2024.114707
Haihong Liu , Huan Jin , Guanyu Xiao , Le wang , Yongsheng Wu , Fang Liu , Chao Zhou , Xiaochuan Liu , Jinggang Qin
The Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) is developing the high-current REBCO cable-in-conduit conductor (CICC) for use in the Central Solenoid (CS) coil of the next generation nuclear fusion device. The aim is to develop a CICC comprising six REBCO sub-cables to satisfy the requirements of operation with a current of around 46 kA and a peak field of up to 20 T. Sub-cables, as crucial components within CICCs, play a pivotal role in ensuring the mechanical support strength and current-carrying stability of the entire CS coil system. Therefore, a process was developed in this paper for the first time to make a sub-cable that meets the requirements by inserting the Highly Flexible REBCO Cable (HFRC) cable into a copper tube and compacting it. Meanwhile, the feasibility and reliability of this process were evaluated and optimized using experimental methods at 77 K and self-field. The result indicated that sub-cables without copper tape protection experienced a 25.42% degradation in critical current (IC). In contrast, when at least one protective layer of copper tape was used as a protective buffer between the HFRC cable and the inner wall of the copper tube, the IC remained stable within a 1% error margin after compaction. These findings demonstrate the feasibility of this sub-cable preparation process. Meanwhile, this compaction method provides a solid process foundation for the development of full-size conductors in future magnet applications.
{"title":"Experimental study on the critical current in highly flexible REBCO cables under copper tube compaction","authors":"Haihong Liu , Huan Jin , Guanyu Xiao , Le wang , Yongsheng Wu , Fang Liu , Chao Zhou , Xiaochuan Liu , Jinggang Qin","doi":"10.1016/j.fusengdes.2024.114707","DOIUrl":"10.1016/j.fusengdes.2024.114707","url":null,"abstract":"<div><div>The Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) is developing the high-current REBCO cable-in-conduit conductor (CICC) for use in the Central Solenoid (CS) coil of the next generation nuclear fusion device. The aim is to develop a CICC comprising six REBCO sub-cables to satisfy the requirements of operation with a current of around 46 kA and a peak field of up to 20 T. Sub-cables, as crucial components within CICCs, play a pivotal role in ensuring the mechanical support strength and current-carrying stability of the entire CS coil system. Therefore, a process was developed in this paper for the first time to make a sub-cable that meets the requirements by inserting the Highly Flexible REBCO Cable (HFRC) cable into a copper tube and compacting it. Meanwhile, the feasibility and reliability of this process were evaluated and optimized using experimental methods at 77 K and self-field. The result indicated that sub-cables without copper tape protection experienced a 25.42% degradation in critical current (<em>I<sub>C</sub></em>). In contrast, when at least one protective layer of copper tape was used as a protective buffer between the HFRC cable and the inner wall of the copper tube, the <em>I<sub>C</sub></em> remained stable within a 1% error margin after compaction. These findings demonstrate the feasibility of this sub-cable preparation process. Meanwhile, this compaction method provides a solid process foundation for the development of full-size conductors in future magnet applications.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114707"},"PeriodicalIF":1.9,"publicationDate":"2024-11-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571253","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-02DOI: 10.1016/j.fusengdes.2024.114705
Chao Li , J. Douglas Way , Thomas F. Fuerst , Colin A. Wolden
Direct internal recycling (DIR) refers to the process of recovering pure hydrogen isotopes (D/T) from helium and other impurities in the fusion plasma exhaust and directing them back to the fuel injection system. Increasing the exhaust fraction purified through DIR significantly reduces the size and cost of the tritium plant and provides additional benefits including reduced requirements for both the tritium startup inventory and tritium breeding ratio. Metal foil pumps (MFPs) are the dominant technology for this separation, relying on the concept of superpermeation. We recently demonstrated that PdCu foils operated at low temperature provide both exceptional flux and resilience to helium absorption as the DIR fraction is increased. Herein we design and demonstrate continuous and semi-batch DIR processes using PdCu MFPs. Under continuous processing, stable performance was observed for DIR fractions up to 92 %. In addition, we demonstrate a semi-batch process capable of extending the DIR fraction to unity (> 99.8 %). Under the experimental conditions described within a PdCu MFP area of ∼22 m2 would be sufficient to process the fusion exhaust with 92 % DIR fraction at expected flowrates of 100 Pa·m3·s−1 for a future fusion power plant.
直接内部循环（DIR）是指从聚变等离子体废气中的氦和其他杂质中回收纯氢同位素（D/T），并将其导回到燃料喷射系统的过程。通过 DIR 增加排气纯化部分可显著降低氚设备的尺寸和成本，并带来其他好处，包括减少对氚启动库存和氚繁殖率的要求。金属箔泵（MFP）是这种分离的主流技术，它依赖于超渗透概念。我们最近演示了在低温下运行的钯铜箔，随着 DIR 分数的增加，可提供卓越的通量和对氦吸收的弹性。在此，我们设计并演示了使用 PdCu MFP 的连续和半批量 DIR 工艺。在连续处理过程中，观察到 DIR 分数高达 92% 时性能稳定。此外，我们还演示了一种半批量工艺，该工艺能够将 DIR 分数提高到统一水平（99.8%）。在所述实验条件下，PdCu MFP 面积为 ∼22 m2，足以处理未来核聚变电厂在预期流量为 100 Pa-m3-s-1 时产生的含 92 % DIR 的核聚变废气。
{"title":"Direct internal recycling fractions approaching unity","authors":"Chao Li , J. Douglas Way , Thomas F. Fuerst , Colin A. Wolden","doi":"10.1016/j.fusengdes.2024.114705","DOIUrl":"10.1016/j.fusengdes.2024.114705","url":null,"abstract":"<div><div>Direct internal recycling (DIR) refers to the process of recovering pure hydrogen isotopes (D/T) from helium and other impurities in the fusion plasma exhaust and directing them back to the fuel injection system. Increasing the exhaust fraction purified through DIR significantly reduces the size and cost of the tritium plant and provides additional benefits including reduced requirements for both the tritium startup inventory and tritium breeding ratio. Metal foil pumps (MFPs) are the dominant technology for this separation, relying on the concept of superpermeation. We recently demonstrated that PdCu foils operated at low temperature provide both exceptional flux and resilience to helium absorption as the DIR fraction is increased. Herein we design and demonstrate continuous and semi-batch DIR processes using PdCu MFPs. Under continuous processing, stable performance was observed for DIR fractions up to 92 %. In addition, we demonstrate a semi-batch process capable of extending the DIR fraction to unity (> 99.8 %). Under the experimental conditions described within a PdCu MFP area of ∼22 m<sup>2</sup> would be sufficient to process the fusion exhaust with 92 % DIR fraction at expected flowrates of 100 Pa·m<sup>3</sup>·<em>s</em><sup>−1</sup> for a future fusion power plant.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114705"},"PeriodicalIF":1.9,"publicationDate":"2024-11-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571252","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-01DOI: 10.1016/j.fusengdes.2024.114688
S. Molokov , G. Politis
Parametric study of main flow characteristics in magnetohydrodynamic flow in the exit duct from a liquid metal blanket has been performed. The flow in such a duct occurs in a nonuniform, decreasing magnetic field. The duct wall is electrically conducting. The wall conductance ratio, , and the gradient of the field, , have been varied in a wide range, and covering most of the cases relevant to fusion. The studies have been performed with asymptotic methods for high values of the Hartmann number and interaction parameter, as well as with FLUENT. Pressure drop correlation has been developed, which is important for blanket design. The results show a significant increase in three-dimensional effects with decreasing wall conductance ratio. For small values of , a stagnant zone is present in the nonuniform field region for all the values of the field gradient, which may create difficulties for tritium removal. If the exit duct is very long, the three-dimensional pressure drop is relatively low, but the effect of the nonuniform magnetic field on the velocity profiles is significant.
对液态金属毯出口管道中磁流体动力流动的主要流动特性进行了参数研究。这种管道中的流动发生在非均匀的递减磁场中。管道壁是导电的。管壁导电率 c 和磁场梯度 γ 的变化范围很广，分别为 0.01≤c≤0.5 和 0.3≤γ≤0.8，涵盖了与核聚变有关的大多数情况。研究采用了哈特曼数和相互作用参数高值的渐近方法以及 FLUENT。还开发了压降相关性，这对橡皮布设计非常重要。结果表明，随着壁导比的减小，三维效应显著增加。在 c 值较小的情况下，所有场梯度值的非均匀场区域都会出现一个停滞区，这可能会给氚的去除造成困难。如果出口管道很长，三维压降相对较低，但非均匀磁场对速度曲线的影响很大。
{"title":"Parametric study of liquid metal flows in conducting circular ducts in a strong nonuniform magnetic field","authors":"S. Molokov , G. Politis","doi":"10.1016/j.fusengdes.2024.114688","DOIUrl":"10.1016/j.fusengdes.2024.114688","url":null,"abstract":"<div><div>Parametric study of main flow characteristics in magnetohydrodynamic flow in the exit duct from a liquid metal blanket has been performed. The flow in such a duct occurs in a nonuniform, decreasing magnetic field. The duct wall is electrically conducting. The wall conductance ratio, <span><math><mi>c</mi></math></span>, and the gradient of the field, <span><math><mi>γ</mi></math></span>, have been varied in a wide range, <span><math><mrow><mn>0</mn><mo>.</mo><mn>01</mn><mo>≤</mo><mi>c</mi><mo>≤</mo><mn>0</mn><mo>.</mo><mn>5</mn></mrow></math></span> and <span><math><mrow><mn>0</mn><mo>.</mo><mn>3</mn><mo>≤</mo><mi>γ</mi><mo>≤</mo><mn>0</mn><mo>.</mo><mn>8</mn></mrow></math></span> covering most of the cases relevant to fusion. The studies have been performed with asymptotic methods for high values of the Hartmann number and interaction parameter, as well as with FLUENT. Pressure drop correlation has been developed, which is important for blanket design. The results show a significant increase in three-dimensional effects with decreasing wall conductance ratio. For small values of <span><math><mi>c</mi></math></span>, a stagnant zone is present in the nonuniform field region for all the values of the field gradient, which may create difficulties for tritium removal. If the exit duct is very long, the three-dimensional pressure drop is relatively low, but the effect of the nonuniform magnetic field on the velocity profiles is significant.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114688"},"PeriodicalIF":1.9,"publicationDate":"2024-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571250","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-01DOI: 10.1016/j.fusengdes.2024.114704
Shahab Ud-Din Khan , Ayesha Alam , Muhammad Abdullah , Ahmad Ali , Ali Hussain , Sehrish Shakir , Shahzaib Zahid , Riaz Khan
The design of a divertor is critically dependent on managing power deposition, erosion effects, and plasma configuration. An upgraded double-null divertor configuration has been developed for the Pakistan Spherical Tokamak (PST), featuring graphite targets that are actively cooled and designed to withstand a peak heat flux of 0.3 MW/m² at a pressure of 0.1 MPa. This paper presents a comprehensive thermal-hydraulic and material analysis for the upgraded flat-type divertor mock-up system, covering aspects such as material surface heat load, peak temperature rise (∆T °C) on the mock-up, and surface temperature increase in the cooling channel. The analysis includes total deformation (mm) and equivalent strain for the inner vertical target (IVT), outer vertical target (OVT), and dome structure, along with material comparison. The recommended SST K-ω turbulence model is utilized in the pressure-based transient analysis, with an inlet velocity of 1.5 m s-1 and an inlet temperature of 16.8 °C. A comparative study of the material and thermal-hydraulic analyses was performed using CFD and RELAP5. The findings reveal that graphite is more suitable than tungsten for the PST's upgraded divertor system, demonstrating its effectiveness as the preferred surface material to address heat enhancement challenges in the PST.
{"title":"Thermal hydraulic and material analysis of upgraded flat-type Graphite divertor mock-up for Pakistan Spherical Tokamak (PST)","authors":"Shahab Ud-Din Khan , Ayesha Alam , Muhammad Abdullah , Ahmad Ali , Ali Hussain , Sehrish Shakir , Shahzaib Zahid , Riaz Khan","doi":"10.1016/j.fusengdes.2024.114704","DOIUrl":"10.1016/j.fusengdes.2024.114704","url":null,"abstract":"<div><div>The design of a divertor is critically dependent on managing power deposition, erosion effects, and plasma configuration. An upgraded double-null divertor configuration has been developed for the Pakistan Spherical Tokamak (PST), featuring graphite targets that are actively cooled and designed to withstand a peak heat flux of 0.3 MW/m² at a pressure of 0.1 MPa. This paper presents a comprehensive thermal-hydraulic and material analysis for the upgraded flat-type divertor mock-up system, covering aspects such as material surface heat load, peak temperature rise (∆T °C) on the mock-up, and surface temperature increase in the cooling channel. The analysis includes total deformation (mm) and equivalent strain for the inner vertical target (IVT), outer vertical target (OVT), and dome structure, along with material comparison. The recommended SST K-ω turbulence model is utilized in the pressure-based transient analysis, with an inlet velocity of 1.5 m s<sup>-1</sup> and an inlet temperature of 16.8 °C. A comparative study of the material and thermal-hydraulic analyses was performed using CFD and RELAP5. The findings reveal that graphite is more suitable than tungsten for the PST's upgraded divertor system, demonstrating its effectiveness as the preferred surface material to address heat enhancement challenges in the PST.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114704"},"PeriodicalIF":1.9,"publicationDate":"2024-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571251","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-31DOI: 10.1016/j.fusengdes.2024.114698
I. Catanzaro , G. Bongiovì , S. Giambrone , A. Gioè , P. Arena , A. Del Nevo , M. Eboli , S. Basile , R. Burlon , P. Chiovaro , P.A. Di Maio , M. Giardina , A. Quartararo , E. Tomarchio , E. Vallone
ENEA, as part of the EUROfusion consortium, is working at the realization of STEAM, an experimental facility integrated into the Water-cooled lithium lead-thermal-HYDRAulic (W-HYDRA) platform, envisaged for the qualification of the EU-DEMO steam generator during the different operational phases of a fusion reactor. It will aim at testing and qualifying the technological solutions to be adopted for the Primary Heat Transport System (PHTS) of the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB). In particular, a mock-up of the Once-Through Steam Generator (OTSG) foreseen for the PHTS of the DEMO WCLL BB will be designed, installed and tested within STEAM. In this regard, a campaign of numerical structural analyses and a preliminary flow-induced vibration study have been carried out at the University of Palermo, in close collaboration with ENEA Brasimone. Firstly, a pipe stress analysis has been conducted focussing on the structural performances of the surge line devoted to connecting the STEAM facility to the pressurizer, looking at the stress and displacement arising under the nominal operating conditions so to propose proper design modifications. Secondly, in order to investigate the structural performances of the DEMO OTSG mock-up in the envisaged nominal operating scenario, a 3D steady-state thermo-mechanical FEM analysis has been carried out. The outcomes have allowed selecting the most critical regions in view of the adopted structural design code. Lastly, in order to preliminarily assess the potential onset of vibration-induced issues within the tubes of the DEMO OTSG mock-up, a preliminary analytical study has been carried out adopting formulae available in literature. The scope has been to establish if vibration-induced issues in the tubes can be reasonably excluded or if they could represent a tangible concern, to be further assessed. Models, assumptions and outcomes are provided and critically discussed.
{"title":"Numerical structural analysis and flow-induced vibration study in support of the design of the EU-DEMO once-through steam generator mock-up for the STEAM experimental facility","authors":"I. Catanzaro , G. Bongiovì , S. Giambrone , A. Gioè , P. Arena , A. Del Nevo , M. Eboli , S. Basile , R. Burlon , P. Chiovaro , P.A. Di Maio , M. Giardina , A. Quartararo , E. Tomarchio , E. Vallone","doi":"10.1016/j.fusengdes.2024.114698","DOIUrl":"10.1016/j.fusengdes.2024.114698","url":null,"abstract":"<div><div>ENEA, as part of the EUROfusion consortium, is working at the realization of STEAM, an experimental facility integrated into the Water-cooled lithium lead-thermal-HYDRAulic (W-HYDRA) platform, envisaged for the qualification of the EU-DEMO steam generator during the different operational phases of a fusion reactor. It will aim at testing and qualifying the technological solutions to be adopted for the Primary Heat Transport System (PHTS) of the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB). In particular, a mock-up of the Once-Through Steam Generator (OTSG) foreseen for the PHTS of the DEMO WCLL BB will be designed, installed and tested within STEAM. In this regard, a campaign of numerical structural analyses and a preliminary flow-induced vibration study have been carried out at the University of Palermo, in close collaboration with ENEA Brasimone. Firstly, a pipe stress analysis has been conducted focussing on the structural performances of the surge line devoted to connecting the STEAM facility to the pressurizer, looking at the stress and displacement arising under the nominal operating conditions so to propose proper design modifications. Secondly, in order to investigate the structural performances of the DEMO OTSG mock-up in the envisaged nominal operating scenario, a 3D steady-state thermo-mechanical FEM analysis has been carried out. The outcomes have allowed selecting the most critical regions in view of the adopted structural design code. Lastly, in order to preliminarily assess the potential onset of vibration-induced issues within the tubes of the DEMO OTSG mock-up, a preliminary analytical study has been carried out adopting formulae available in literature. The scope has been to establish if vibration-induced issues in the tubes can be reasonably excluded or if they could represent a tangible concern, to be further assessed. Models, assumptions and outcomes are provided and critically discussed.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"209 ","pages":"Article 114698"},"PeriodicalIF":1.9,"publicationDate":"2024-10-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142561282","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-29DOI: 10.1016/j.fusengdes.2024.114696
Y.X. Zhu , Q. Zang , W. Chu , M․Yu Kantor , G.L. Lin , M.F. Ren
This paper presents the latest development of a tangential TVTS (Television Thomson scattering) system on the EAST Tokamak and the preliminary results obtained recently. For the goals of the thousand-second H-mode on EAST, it is obviously essential for the experimental operation and physical research to develop a high spatial resolution edge Thomson scattering diagnostic system. The new system uses a tangential optical path, mainly covering plasma ρ = 0.8 – 1.1 region. A high energy 532 nm laser was designed as the light source, which can realize the stable output of 10 Hz/3.5 J in the time of a thousand seconds through the new optical structure technology. In order to improve the scattering light splitting efficiency, the transmission grating is used as the main splitting device. We use a three-step image enhancement amplification technique so that the detection efficiency of the system can reach 3000 counts/photon. At present, the spatial resolution of the system is about 3 mm, and the measurement electron temperature (Te) range is 50 eV - 2 keV, the electron temperature error is expected <7 %.
{"title":"Preliminary results and analysis of a tangential TV Thomson scattering diagnostic system on EAST","authors":"Y.X. Zhu , Q. Zang , W. Chu , M․Yu Kantor , G.L. Lin , M.F. Ren","doi":"10.1016/j.fusengdes.2024.114696","DOIUrl":"10.1016/j.fusengdes.2024.114696","url":null,"abstract":"<div><div>This paper presents the latest development of a tangential TVTS (Television Thomson scattering) system on the EAST Tokamak and the preliminary results obtained recently. For the goals of the thousand-second H-mode on EAST, it is obviously essential for the experimental operation and physical research to develop a high spatial resolution edge Thomson scattering diagnostic system. The new system uses a tangential optical path, mainly covering plasma ρ = 0.8 – 1.1 region. A high energy 532 nm laser was designed as the light source, which can realize the stable output of 10 Hz/3.5 <em>J</em> in the time of a thousand seconds through the new optical structure technology. In order to improve the scattering light splitting efficiency, the transmission grating is used as the main splitting device. We use a three-step image enhancement amplification technique so that the detection efficiency of the system can reach 3000 counts/photon. At present, the spatial resolution of the system is about 3 mm, and the measurement electron temperature (T<sub>e</sub>) range is 50 eV - 2 keV, the electron temperature error is expected <7 %.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"208 ","pages":"Article 114696"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537947","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-29DOI: 10.1016/j.fusengdes.2024.114700
Chenxi Hu , Shanliang Zheng , Yuanjie Li
In the China Fusion Engineering Test Reactor (CFETR), the occurrence of a Loss of Vacuum Accident (LOVA) could result in high-velocity airflow entering the vacuum vessel, causing plasma disruption and potentially posing a significant safety hazard. In this study, the LOVA involving multiple small and minor breaches in the vacuum vessel (VV) of the CFETR is analyzed through modeling and simulation utilizing the ANSYS Fluent software. The study examines the influence of the initial pressure in VV, the total area, number, position, and distribution of breaking points on gas flow characteristics during the LOVA, based on the validation of the numerical model. The analysis includes the examination of the evolution and disappearance of the Mach disk resulting from supersonic fluid under the conditions of LOVA. With a constant total area, an increase in the number of distinct breaking points would result in a reduction of the critical mass flow rate of the jet. This decrease would decelerate the decline of the flow rate, consequently extending the necessary stability time. Moreover, the findings indicate that as the number of breaking points increases, the size of the Mach disk observed along the central flow axis diminishes, while the numerical value demonstrates a rising pattern.
{"title":"Parametric study of the breaking scenarios during the loss of vacuum accident for CFETR","authors":"Chenxi Hu , Shanliang Zheng , Yuanjie Li","doi":"10.1016/j.fusengdes.2024.114700","DOIUrl":"10.1016/j.fusengdes.2024.114700","url":null,"abstract":"<div><div>In the China Fusion Engineering Test Reactor (CFETR), the occurrence of a Loss of Vacuum Accident (LOVA) could result in high-velocity airflow entering the vacuum vessel, causing plasma disruption and potentially posing a significant safety hazard. In this study, the LOVA involving multiple small and minor breaches in the vacuum vessel (VV) of the CFETR is analyzed through modeling and simulation utilizing the ANSYS Fluent software. The study examines the influence of the initial pressure in VV, the total area, number, position, and distribution of breaking points on gas flow characteristics during the LOVA, based on the validation of the numerical model. The analysis includes the examination of the evolution and disappearance of the Mach disk resulting from supersonic fluid under the conditions of LOVA. With a constant total area, an increase in the number of distinct breaking points would result in a reduction of the critical mass flow rate of the jet. This decrease would decelerate the decline of the flow rate, consequently extending the necessary stability time. Moreover, the findings indicate that as the number of breaking points increases, the size of the Mach disk observed along the central flow axis diminishes, while the numerical value demonstrates a rising pattern.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"208 ","pages":"Article 114700"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537951","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-29DOI: 10.1016/j.fusengdes.2024.114702
A. Wang , T.F. Sun , W. Chen , B.T. Cui , J.M. Gao , S.Y. Liang , M.Y. He , X.Q. Ji
This paper details the design, construction, and analysis of resonant magnetic perturbation (RMP) with 16 in-vessel coils on the HL-3 to investigate the interactions between resonant magnetic perturbations and magnetohydrodynamic instabilities. The toroidal × poloidal = 8 × 2 arrangement allows for a more flexible magnetic field and spectrum shape. The RMP coil was designed as a four-turn double-layer winding. The coils are adopted with a water-cooled hollow copper conductor insulated with a MgO layer and then housed inside a welded stainless steel shell. The thermal and electromagnetic loads of the coil were evaluated by simulation, which meets the requirements. Four power supplies are integrated into the HL-3 system that can provide DC and AC with various waveforms. The maximum DC amplitude is 3 kA, while the AC frequency can reach up to 1 kHz. In the HL-3 H-mode plasmas, an experimental result of edge localized mode (ELM) mitigation with n = 1 RMPs which are oddly paired by 8 RMP coils is also given for the first time in this paper.
{"title":"Final development and preliminary experiment progress of in-vessel resonant magnetic perturbation coils system on HL-3 tokamak","authors":"A. Wang , T.F. Sun , W. Chen , B.T. Cui , J.M. Gao , S.Y. Liang , M.Y. He , X.Q. Ji","doi":"10.1016/j.fusengdes.2024.114702","DOIUrl":"10.1016/j.fusengdes.2024.114702","url":null,"abstract":"<div><div>This paper details the design, construction, and analysis of resonant magnetic perturbation (RMP) with 16 in-vessel coils on the HL-3 to investigate the interactions between resonant magnetic perturbations and magnetohydrodynamic instabilities. The toroidal × poloidal = 8 × 2 arrangement allows for a more flexible magnetic field and spectrum shape. The RMP coil was designed as a four-turn double-layer winding. The coils are adopted with a water-cooled hollow copper conductor insulated with a MgO layer and then housed inside a welded stainless steel shell. The thermal and electromagnetic loads of the coil were evaluated by simulation, which meets the requirements. Four power supplies are integrated into the HL-3 system that can provide DC and AC with various waveforms. The maximum DC amplitude is 3 kA, while the AC frequency can reach up to 1 kHz. In the HL-3 H-mode plasmas, an experimental result of edge localized mode (ELM) mitigation with <em>n</em> = 1 RMPs which are oddly paired by 8 RMP coils is also given for the first time in this paper.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"208 ","pages":"Article 114702"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537952","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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