Fusion Engineering and Design最新文献

英文中文

A risk analysis for Helium Cooling System (HCS) of HCCP-TBS

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-06 DOI: 10.1016/j.fusengdes.2025.114921

Daniel Moya Alonso , Youngmin Lee , Enric Bargalló Font , Chang Shuk Kim , Myungho Kim , Changwook Shin , Seok-Kwon Son , Mu-Young Ahn , Martín Gorriarán Vicario , Javier Leboreiro Iriarte

Climate change is a reality. Every day, the effects of it are visible and one of the main challenges is to achieve carbon neutrality by 2050. With this goal, one increasing technology in clean energy generation is nuclear fusion. Nuclear fusion needs unique working conditions, such as: extremely elevated temperatures, strong electromagnetic fields, and ultra-high vacuum conditions. But another challenging milestone should be taken into account: breeding tritium. Tritium quantity is naturally scarce on Earth and has to be generated, which is costly and currently very limited. One of the missions for the later stages of ITER operation is to demonstrate the feasibility of producing tritium from lithium within the vacuum vessel, making breeding blankets a key component of ITER.

Given its relevance, there are several proposals to breed tritium, from using helium or water as a coolant, or liquid types like lithium-lead (LiPb) alloy or solid types such as lithiated ceramic pebbles as breeder. One of the proposals being currently developed, is the Helium Cooled Ceramic Pebble (HCCP), where lithiated ceramic pebbles are used as breeder and helium, as coolant. HCCP is designed to test tritium breeding, recovery, and withstand reactor loads. A key function is heating removal, with the Helium Cooling System (HCS) essential for ITER's safety and availability.

This paper describes the Reliability, Availability, Maintainability and Inspectability (RAMI) analysis and preliminary risk assessment (HAZOP) performed, together with the proposed design, operation and maintenance suggestions needed to achieve the mail goal: a reliable and safe operation for HCCP-HCS.

{"title":"A risk analysis for Helium Cooling System (HCS) of HCCP-TBS","authors":"Daniel Moya Alonso , Youngmin Lee , Enric Bargalló Font , Chang Shuk Kim , Myungho Kim , Changwook Shin , Seok-Kwon Son , Mu-Young Ahn , Martín Gorriarán Vicario , Javier Leboreiro Iriarte","doi":"10.1016/j.fusengdes.2025.114921","DOIUrl":"10.1016/j.fusengdes.2025.114921","url":null,"abstract":"<div><div>Climate change is a reality. Every day, the effects of it are visible and one of the main challenges is to achieve carbon neutrality by 2050. With this goal, one increasing technology in clean energy generation is nuclear fusion. Nuclear fusion needs unique working conditions, such as: extremely elevated temperatures, strong electromagnetic fields, and ultra-high vacuum conditions. But another challenging milestone should be taken into account: breeding tritium. Tritium quantity is naturally scarce on Earth and has to be generated, which is costly and currently very limited. One of the missions for the later stages of ITER operation is to demonstrate the feasibility of producing tritium from lithium within the vacuum vessel, making breeding blankets a key component of ITER.</div><div>Given its relevance, there are several proposals to breed tritium, from using helium or water as a coolant, or liquid types like lithium-lead (LiPb) alloy or solid types such as lithiated ceramic pebbles as breeder. One of the proposals being currently developed, is the Helium Cooled Ceramic Pebble (HCCP), where lithiated ceramic pebbles are used as breeder and helium, as coolant. HCCP is designed to test tritium breeding, recovery, and withstand reactor loads. A key function is heating removal, with the Helium Cooling System (HCS) essential for ITER's safety and availability.</div><div>This paper describes the Reliability, Availability, Maintainability and Inspectability (RAMI) analysis and preliminary risk assessment (HAZOP) performed, together with the proposed design, operation and maintenance suggestions needed to achieve the mail goal: a reliable and safe operation for HCCP-HCS.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114921"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143552887","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}

引用次数: 0

Preliminary fault analysis of EFCC system in the Divertor Tokamak Test (DTT) facility

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-06 DOI: 10.1016/j.fusengdes.2025.114915

Antonio Iaiunese , Raffaele Albanese , Andrea Gaetano Chiariello , Pasquale Zumbolo , Raffaele Martone

Tokamaks use Error Field Correction Coils to reduce the error fields due to misalignments and assembly tolerances in the main magnetic system, with the aim of avoiding plasma instabilities. These coils may fail for various reasons, e.g., problems in insulation or power supply. This paper presents a method to evaluate the capability of correcting error fields in case of failure of one or more coils. The effectiveness is assessed in terms of the Three Mode Error Index metric. The paper shows the results obtained for the Divertor Tokamak Test facility, but the method is quite general and can be applied to any device.

引用次数: 0

Development of transport calculation for unstructured mesh geometry in cosRMC

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-06 DOI: 10.1016/j.fusengdes.2025.114933

Zhenyu Wang, Shichang Liu, Yixue Chen

The traditional constructive solid geometry (CSG) modeling approach fails to meet the requirements of complex geometric transport, fusion neutronics analysis, multi-physics coupling calculations, and large-scale nuclear radiation effect simulations. To tackle these challenges, this study developed unstructured mesh (UM) transport and statistical functionalities in the Monte Carlo particle transport code, cosRMC, based on the MOAB and libMesh libraries. These functionalities allow for accurate particle positioning and tracking within UM elements, and the transport process is accelerated through a neighbor element acceleration technique. The track-length estimator is then used for tally results. To verify the computational accuracy and efficiency of the UM feature, three test cases were constructed and analyzed: a box case, a fuel rod case, and a tokamak case. The computational results demonstrate that the k_eff values calculated by UM-cosRMC geometry transport function align well with those from traditional CSG and DAGMC geometries, and the UM tally results are consistent with those of UM-MCNP calculations. In terms of computational efficiency, after employing the neighbor mesh acceleration technique, the speed of UM geometry calculations improved significantly. The UM geometry calculation efficiency based on the libMesh library is higher than that based on the MOAB library, with the former taking 1.18 to 10.42 times longer than CSG geometry calculations and 0.37 to 0.96 times that of DAGMC geometry. Compared to UM-MCNP calculation efficiency, UM-cosRMC shows a distinct advantage, reducing computation time by >70 % with the same mesh. Additionally, this study analyzed the computational efficiency of UM transport based on both libraries, and the results indicate that the libMesh-based UM calculations outperform in terms of UM reading speed, memory usage, and transport computation speed.

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引用次数: 0

Thermofluid-dynamic assessment of the dual cooling scheme EU-DEMO divertor cassette

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.fusengdes.2025.114903

F.M. Castrovinci , A. Quartararo , S. Basile , G. Bongiovì , R. Burlon , P. Chiovaro , P.A. Di Maio , A. Gioè , S. Maggio , G. Mazzone , E. Tomarchio , S. Vacca , E. Vallone , J.-H. You

The EU-DEMO divertor is currently foreseen to employ a dual cooling scheme, which involves two separate cooling circuits for the Cassette Body (CB) and the Vertical Targets (VTs), provided with cooling water at distinct operating conditions. Following the EU-DEMO project Gate Review 1, the cassette design has been revised to employ a high-temperature (295-328 °C) and high-pressure (15.5 MPa) coolant for the CB cooling circuit. These design assumptions present novel challenges in achieving uniform and effective cooling of the structure, to ensure reliable operation over the intended lifetime. Furthermore, alternative materials for the VTs-supporting Target Support Bodies (TSBs), namely CuCrZr, SS 316Ti and Eurofer97, are currently being evaluated to extend the operational lifetime of these components under irradiation. The University of Palermo, in collaboration with EUROfusion and ENEA, analysed the steady-state thermal-hydraulic performance of the EU-DEMO divertor cooling circuits with an integrated conjugate heat transfer analysis campaign of the entire divertor cassette, with the twofold aim to evaluate the CB thermal performance under the revised coolant conditions and to compare the temperature distribution in the TSBs with the different materials considered. The work was carried out following a numerical approach based on the finite volume method and adopting the ANSYS CFX computational fluid-dynamic commercial code. Models, loads and boundary conditions are herewith reported and critically discussed, together with the main outcomes obtained.

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引用次数: 0

Steady-state Superconducting Advanced Spherical Tokamak Reactor (SASTR)

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.fusengdes.2025.114900

Yoshio Nagayama , Takaaki Fujita

A concept for a Steady-state Superconducting Advanced Spherical Tokamak Reactor (SASTR) with a slim-CS is proposed. This is a spherical tokamak (ST) with an internal transport barrier (ITB) and superconducting magnets. The ITB allows for sufficient bootstrap (BS) current to sustain the plasma. The feasibility of a self-sustained SASTR is investigated by using a set of plasma burning equations. We find that the toroidal magnetic field and the confinement enhancement factor are key parameters to meet the beta limit and density limit criteria, respectively. We estimate the dependence of the cost of electricity (COE) on the aspect ratio and reactor size. We present a conceptual design of a 0.8 GWe SASTR power reactor with a major radius of 4.5 m, an aspect ratio of 1.8, a toroidal field of 2.67 T generated by Nb₃Sn superconducting magnets, and a plasma current of 24.5 MA driven fully by the BS current.

引用次数: 0

Insight into the structural integrity assessment of the European DEMO Fusion Reactor Divertor (Part I: General discussion)

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.fusengdes.2025.114928

M. Muscat , P. Mollicone , N. Mantel , J.H. You

This paper gives a description of the conceptual design of the DEMO divertor highlighting its two main functions and the main loading conditions acting on it during operation. The paper stresses the importance of Design by analysis (DBA) and structural integrity assessments at the conceptual design phase. It explains why currently the predominant DBA approach taken by the DEMO divertor group within the EUROfusion consortium is the elastic approach rather than the elastoplastic approach. Background to the DBA elastic approach describing linearization of stresses through the wall thickness and to stress categorisation that allows different stress allowables to be used for a more efficient design is given. Throughout the paper, the RCC MRx code structural integrity P Type and S Type checks as applied to the DEMO divertor components are summarised and difficulties in applying the elastic design rules to the divertor model are highlighted. These difficulties range from issues of code interpretation as applied to the DEMO divertor geometry to limited availability of material properties for Eurofer97 steel (a low activation steel) which is the main material selected for the divertor. In a follow up paper (Part II: Application Examples), some of these difficulties are demonstrated by considering two stress classification lines in the model of the current DEMO divertor model.

In this paper, the phenomenon of elastic follow up is explained and DEMO divertor load cases when this has to be taken into consideration are discussed. A short explanation on Neuber's rule, which is used to calculate estimates of plastic strain and to differentiate between primary and secondary stresses for Type S checks is also given. At the end of the paper, the conclusions summarise the interpretation and issues encountered in the application of the RCC-MRx rules for a typical nuclear fusion reactor component such as the DEMO divertor design.

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引用次数: 0

Development of a real-time tangential dispersion interferometer system and its first results for density feedback on EAST

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.fusengdes.2025.114930

Yuyang Liu , Weiming Li , Yao Zhang , Shouxin Wang , Jiamin Zhang , Hui Lian , Bo Hong , Wenmin Zhang , Yuan Yao , Yinxian Jie , Haiqing Liu

A horizontal tangential dispersion interferometer (TDI) system has been developed for the Experimental Advanced Superconducting Tokamak (EAST) to enhance the accuracy of density feedback. This system offers several advantages, including vibration insensitivity and a reduced risk of fringe jump errors compared to far-infrared interferometers, making it more suitable for density feedback control.

The TDI system uses two horizontal windows for tangential density measurements. Experimental results demonstrate that the TDI provides stable signals and enhances overall system reliability. Statistical analysis demonstrates that the TDI data have a measurement success rate of approximately 96%. Moreover, the temporal resolution of the system, at 20 µs, meets the requirements for real-time feedback control on EAST. The system employs a modified density real-time processing system and has been evaluated for density feedback control. During density feedback control experiments, the TDI system demonstrated notable advantages and stability, making it more suitable for future high-density burning plasma devices.

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引用次数: 0

Progress of “Plasma and Fusion Cloud” research data platform towards “Open Science”

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-04 DOI: 10.1016/j.fusengdes.2025.114898

Hideya Nakanishi , Masahiko Emoto , Arimichi Takayama , Takashi Yamamoto , Kenjiro Yamanaka , Shigeo Urushidani , Masaki Ohsuna , Setsuo Imazu , Masanobu Yoshida , Miki Nonomura , Shinsuke Satake , Yasushi Todo , Masaki Osakabe , Ryuichi Sakamoto

A new cloud platform to realize the plasma and fusion experimental data ecosystem, named the “Plasma and Fusion Cloud,” has been technically verified on some fundamental issues. Such an enormous amount of diagnostic data requires a high-performance computing (HPC) platform not only for LHD physics data analyses but also for next-generation experiments, such as ITER and JT-60SA. Performance evaluation has been done at NIFS by using the HPC supercomputer “Raijin” and the LHD primary data storage system, both of which are directly connected by the 100 Gbps Ethernet optical link. The test results show that almost a full bandwidth can be used by means of multiple parallel streams.

In order to make plasma and fusion diagnostic data “FAIR”, all the LHD’s diagnostic data objects are now under way to be registered with digital object identifiers (DOI) for each acquisition node and plasma pulse. In 2023, more than 1.2 million DOIs were issued for the LHD diagnostic data. Cloud technology is also very promising as a high-performance data computing platform, not only for physics data analyses but also for real-time plasma and plant controls. AWS (Amazon Web Service) S3 cloud storage has accepted a proposal to store all 2.0 peta-byte of compressed LHD physics data for open access, under the AWS Open Data Sponsorship Program (ODP). AWS is also one of the commercial providers of computing clouds in the framework of NII’s Research Data Cloud (RDC) in Japan, allowing LHD data users to increase or decrease the computing power they need on demand.

The demonstrations and technical verifications done in this study suggest that a next-generation fusion data research center could be based on cloud technology.

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引用次数: 0

Design of a 170GHz quasi-optical notch filter for microwave-based diagnostics protection on the burning plasma device 设计用于燃烧等离子体装置微波诊断保护的 170GHz 准光学陷波滤波器

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-04 DOI: 10.1016/j.fusengdes.2025.114925

Chen Luo , Xiaopin Tang , Calvin Domier , Xiaoliang Li , Ang Ti , Pengjun Sun , Shasha Qiu , Xinhang Xu , Xianzi Liu , Yilun Zhu , Neville Luhmann

170 GHz gyrotrons will be used for Electron Cyclotron Heating (ECH) on the under-development burning plasma device, with a maximum power output of 16 MW with center frequency of 170 GHz. The reliable 170 GHz quasi-optical notch filters protect microwave-based diagnostic instruments from unabsorbed ECH power. This design utilizes a metal circular periodic structure, which is straightforward to manufacture, to achieve reflection and absorption of 170 GHz electromagnetic waves transmitted through space, thereby achieving the notch filtering effect (attenuation > 60 dB). Unlike the rectangular ring design used in low-frequency quasi-optical notch filters, the circular design caters to the shorter wavelength of 170 GHz, which necessitates reduced fabrication error tolerance. To achieve both high quality filter and higher processing yields, we compensate for the performance degradation of the notch filter caused by processing errors by optimizing the controllable metal circle radius and incident angle. This paper discusses the impact of material size selection and machining accuracy deviations on the performance of the notch filter.

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引用次数: 0

Thermal-hydraulic scaling of the prototypical mock-up for European DEMO HCPB breeding blanket

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-04 DOI: 10.1016/j.fusengdes.2025.114924

Gaurav Verma, Guangming Zhou, Francisco A. Hernández

The Helium-Cooled Pebble Bed (HCPB) blanket concept is a leading candidate for the driver blanket in the European Fusion Demonstration Power Plant (DEMO), developed within the EUROfusion framework's Work Package Breeding Blanket (WPBB). A crucial component of the HCPB blanket is the First Wall (FW), which must withstand high heat fluxes from the plasma while maintaining a uniform temperature distribution as much as possible. This uniformity is achieved through an alternating coolant flow design. The present study focuses on the thermal-hydraulic scaling methodology for the First Wall-Prototype Mock-Up (FW-PMU) to replicate the HCPB blanket's thermal and hydraulic performance. The FW-PMU is designed to emulate the flow distribution and thermal behavior of the HCPB Breeding Blanket (BB) First Wall, including flow patterns from the Breeder Zone Manifold to the fuel-breeder pins. In the present work, a power-to-volume scaling approach is employed to maintain the characteristic time ratio by analyzing non-dimensional parameters, while scaling of component dimensions in the breeder zone manifold ensures a consistent velocity field. The FW-PMU is set for testing in HELOKA, a high-pressure, high-temperature helium facility at the Karlsruhe Institute of Technology to validate the HCPB First Wall design under realistic operating conditions, advancing its potential deployment in fusion energy systems.

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