Kwanwoo Nam , Dongkwon Kang , Moohyun You , Myeongjun Kim , Sang-chul Kim , Chang Hyun Noh , Hyunsoo Kim , Kijung Jung
{"title":"Study on the new manufacture of ITER vacuum vessel thermal shield","authors":"Kwanwoo Nam , Dongkwon Kang , Moohyun You , Myeongjun Kim , Sang-chul Kim , Chang Hyun Noh , Hyunsoo Kim , Kijung Jung","doi":"10.1016/j.fusengdes.2025.114964","DOIUrl":null,"url":null,"abstract":"<div><div>This paper describes the preparation and the status for the manufacturing of the new Vacuum Vessel Thermal Shield (VVTS) for ITER machine. As the delivered VVTS had the corrosion problem at the pipe weld, the VVTS was determined to be repaired and re-manufactured. The new VVTS has no silver coating and its pipe material is changed to exclude corrosion cracking at the cooling pipe. Roughness of the panel surface is reduced to have lower emissivity without silver coating. Several technical efforts to improve the pipe welding are presented in this paper. Stitch weld between the pipe and the panel is carefully tested with various specimens as Pre-Production Sample (PPS). Tack weld position is intended not to be overlapped with the stitch weld. Back shielding at the opposite side of the stitch weld is also applied to minimize the oxidation. Down-slope current of the welding machine is optimized for the pipe stitch weld. Heat correction method after the welding is investigated to confirm the soundness of the weld joint and the base metal. Impact test and corrosion test are performed for the specimens after the heat correction. All the manufactured VVTS panels are subjected to be leak-tested in a huge vacuum chamber with the pressurized cooling pipe with helium gas. Delivery status of the re-manufactured VVTS panels is presented and the anticipated manufacturing schedule are introduced in this paper.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114964"},"PeriodicalIF":1.9000,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Fusion Engineering and Design","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0920379625001644","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
引用次数: 0
Abstract
This paper describes the preparation and the status for the manufacturing of the new Vacuum Vessel Thermal Shield (VVTS) for ITER machine. As the delivered VVTS had the corrosion problem at the pipe weld, the VVTS was determined to be repaired and re-manufactured. The new VVTS has no silver coating and its pipe material is changed to exclude corrosion cracking at the cooling pipe. Roughness of the panel surface is reduced to have lower emissivity without silver coating. Several technical efforts to improve the pipe welding are presented in this paper. Stitch weld between the pipe and the panel is carefully tested with various specimens as Pre-Production Sample (PPS). Tack weld position is intended not to be overlapped with the stitch weld. Back shielding at the opposite side of the stitch weld is also applied to minimize the oxidation. Down-slope current of the welding machine is optimized for the pipe stitch weld. Heat correction method after the welding is investigated to confirm the soundness of the weld joint and the base metal. Impact test and corrosion test are performed for the specimens after the heat correction. All the manufactured VVTS panels are subjected to be leak-tested in a huge vacuum chamber with the pressurized cooling pipe with helium gas. Delivery status of the re-manufactured VVTS panels is presented and the anticipated manufacturing schedule are introduced in this paper.
期刊介绍:
The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.